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At least 955 records · Page 53
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DOE OSTI2014

Quantifying construction and demolition waste: An analytical review

Highlights: • Prevailing C and D waste quantification methodologies are identified and compared. • One specific methodology cannot fulfill all waste quantification scenarios. • A relevance tree for appropriate quantification methodology selection is proposed. • More attentions should be paid to civil and infrastructural works. • Classified information is suggested for making an effective waste management plan. - Abstract: Quantifying construction and demolition (C and D) waste generation is regarded as a prerequisite for the implementation of successful waste management. In literature, various methods have been employed to quantify the C and D waste generation at both regional and project levels. However, an integrated review that systemically describes and analyses all the existing methods has yet to be conducted. To bridge this research gap, an analytical review is conducted. Fifty-seven papers are retrieved based on a set of rigorous procedures. The characteristics of the selected papers are classified according to the following criteria - waste generation activity, estimation level and quantification methodology. Six categories of existing C and D waste quantification methodologies are identified, including site visit method, waste generation rate method, lifetime analysis method, classification system accumulation method, variables modelling method and other particular methods. A critical comparison of the identified methods is given according to their characteristics and implementation constraints. Moreover, a decision tree is proposed for aiding the selection of the most appropriate quantification method in different scenarios. Based on the analytical review, limitations of previous studies and recommendations of potential future research directions are further suggested.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Battery collection in municipal waste management in Japan: Challenges for hazardous substance control and safety

Highlights: • Consumers need to pay attention to the specific collection rules for each type of battery in each municipality in Japan. • 6–10% of zinc carbon and alkaline batteries discarded in Japan currently could be regarded as containing mercury. • Despite announcements by producers and municipalities, only 2.0% of discarded cylindrical dry batteries were insulated. • Batteries made up an average of 4.6% of the total collected small WEEE under the small WEEE recycling scheme in Japan. • Exchangeable batteries were used in almost all of mobile phones, but the removal rate was as low as 22% for mobile phones. - Abstract: To clarify current collection rules of waste batteries in municipal waste management in Japan and to examine future challenges for hazardous substance control and safety, we reviewed collection rules of waste batteries in the Tokyo Metropolitan Area. We also conducted a field survey of waste batteries collected at various battery and small waste electric and electronic equipment (WEEE) collection sites in Tokyo. The different types of batteries are not collected in a uniform way in the Tokyo area, so consumers need to pay attention to the specific collection rules for each type of battery in each municipality. In areas where small WEEE recycling schemes are being operated after the enforcement of the Act on Promotion of Recycling of Small Waste Electrical and Electronic Equipment in Japan in 2013, consumers may be confused about the need for separating batteries from small WEEE (especially mobile phones). Our field survey of collected waste batteries indicated that 6–10% of zinc carbon and alkaline batteries discarded in Japan currently could be regarded as containing mercury. More than 26% of zinc carbon dry batteries currently being discarded may have a lead content above the labelling threshold of the EU Batteries Directive (2006/66/EC). In terms of safety, despite announcements by producers and municipalities about using insulation (tape) on waste batteries to prevent fires, only 2.0% of discarded cylindrical dry batteries were insulated. Our field study of small WEEE showed that batteries made up an average of 4.6% of the total collected small WEEE on a weight basis. Exchangeable batteries were used in almost all of mobile phones, digital cameras, radios, and remote controls, but the removal rate was as low as 22% for mobile phones. Given the safety issues and the rapid changes occurring with mobile phones or other types of small WEEE, discussion is needed among stakeholders to determine how to safely collect and recycle WEEE and waste batteries.

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DOE OSTI2015

The effect of gender and age structure on municipal waste generation in Poland

Highlights: • An effect of gender and age structure on municipal waste generation was presented. • The waste accumulation index is influenced by a number of unemployed women. • Greater share of women in society contributes to greater waste production. • A model describing the analyzed dependences was determined. - Abstract: In this study the effect of gender and age structure on municipal waste generation was investigated. The data from 10-year period, from 2001 to 2010 year, were taken into consideration. The following parameters of gender and age structure were analyzed: men and woman quantity, female to male ratio, number of working, pre-working and post-working age men/women, number of unemployed men/women. The results have showed a strong correlation of annual per capita waste generation rate with number of unemployed women (r = 0.70) and female to male ratio (r = 0.81). This indicates that waste generation rate is more depended on ratio of men and women that on quantitative size of each group. Using the regression analysis a model describing the dependence between female to male ratio, number of unemployed woman and waste quantity was determined. The model explains 70% of waste quantity variation. Obtained results can be used both to improve waste management and to a fuller understanding of gender behavior.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Characterization of radioactive wastes with respect to harmful materials

In addendum 4 to the license of the German KONRAD repository, which considers mainly radiological aspects, a water law permit was issued in order to prevent the pollution of the near-surface groundwater. The water law permit stipulates limitations for 10 radionuclides and 2 groups of radionuclides as well as mass limitations for 94 substances and materials relevant for water protection issues. Two collateral clauses, i.e. additional requirements imposed by the licensing authority, include demands on the monitoring, registering and balancing of non-radioactive harmful substances and materials /1/. In order to fulfill the requirements of the water law permit the German Federal Office for Radiation Protection (BfS) being the operator of the KONRAD repository has developed a concept, which ensures the compliance with all requirements of the water law permit and which provides standardized easy manageable guidance for the waste producers to describe their wastes. On 15 March 2011 the competent water authority, the 'Niedersaechsischer Landesbetrieb fuer Wasserwirtschaft, Kuesten- und Naturschutz' (NLWKN) issued the approval for this concept. Being the most essential part of this concept the procedural method and the developed description of nonradioactive waste package constituents by use of standardized lists of materials and containers is addressed and presented in this paper. The waste producer has to describe his waste package in a standardized way on the base of the lists of materials and containers. For each material in the list a comprehensive description is given comprising the composition, scope of application, quality control measures, thresholds and other data. Each entry in the list has to be approved by NLWKN. The scope of the lists is defined by the waste producers' needs. Using some particular materials as examples, the approval procedure for including materials in the list is described. The procedure of describing the material composition has to be considered in the KONRAD waste acceptance requirements. The respective part of these requirements will be introduced. In order to clarify the procedure of describing waste packages by use of the standardized lists of materials and containers some examples of typical waste package descriptions will be presented. (authors)

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DOE OSTI2013

Incineration of European non-nuclear radioactive waste in the USA

Incineration of dry low level radioactive waste from nuclear stations is a well established process achieving high volume reduction factors to minimise disposal costs and to stabilise residues for disposal. Incineration has also been applied successfully in many European Union member countries to wastes arising from use of radionuclides in medicine, nonnuclear research and industry. However, some nations have preferred to accumulate wastes over many years in decay stores to reduce the radioactive burden at point of processing. After decay and sorting the waste, they then require a safe, industrial scale and affordable processing solution for the large volumes accumulated. This paper reports the regulatory, logistical and technical issues encountered in a programme delivered for Eckert and Ziegler Nuclitec to incinerate safely 100 te of waste collected originally from German research, hospital and industrial centres, applying for the first time a 'burn and return' process model for European waste in the US. The EnergySolutions incinerators at Bear Creek, Oak Ridge, Tennessee, USA routinely incinerate waste arising from the non-nuclear user community. To address the requirement from Germany, EnergySolutions had to run a dedicated campaign to reduce cross-contamination with non-German radionuclides to the practical minimum. The waste itself had to be sampled in a carefully controlled programme to ensure the exacting standards of Bear Creek's license and US emissions laws were maintained. Innovation was required in packaging of the waste to minimise transportation costs, including sea freight. The incineration was inspected on behalf of the German regulator (the BfS) to ensure suitability for return to Germany and disposal. This first 'burn and return' programme has safely completed the incineration phase in February and the arising ash will be returned to Germany presently. The paper reports the main findings and lessons learned on this first of its kind project. (authors)

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DOE OSTI2015

Supplemental Immobilization of Hanford Low-Activity Waste: Cast Stone Augmented Formulation Matrix Tests

More than 56 million gallons of radioactive and hazardous waste are stored in 177 underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in Washington State. The HLW will be vitrified in the HLW facility for ultimate disposal at an offsite federal repository. A portion (~35%) of the LAW will be vitrified in the LAW vitrification facility for disposal onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize all of the wastes destined for those facilities. However, a second facility will be needed for the expected volume of LAW requiring immobilization. Cast Stone, a cementitious waste form, is being considered to provide the required additional LAW immobilization capacity. The Cast Stone waste form must be acceptable for disposal in the IDF. The Cast Stone waste form and immobilization process must be tested to demonstrate that the final Cast Stone waste form can comply with the waste acceptance criteria for the disposal facility and that the immobilization processes can be controlled to consistently provide an acceptable waste form product. A testing program was developed in fiscal year (FY) 2012 describing in detail the work needed to develop and qualify Cast Stone as a waste form for the solidification of Hanford LAW. A statistically designed test matrix was used to evaluate the effects of key parameters on the properties of the Cast Stone as it is initially prepared and after curing. For the processing properties, the water-to-dry-blend mix ratio was the most significant parameter in affecting the range of values observed for each property. The single shell tank (SST) Blend simulant also showed differences in measured properties compared to the other three simulants tested. A review of the testing matrix and results indicated that an additional set of tests would be beneficial to improve the understanding of the impacts noted in the Screening Matrix tests. A set of Cast Stone formulations were devised to augment the original screening test matrix and focus on the range of the test conditions. Fly ash and blast furnace slag were limited to either northwest or southeast and the salt solutions were narrowed to the Average and the SST Blend at the 7.8M Na concentration. To fill in the matrix, a mix ratio of 0.5 was added. In addition, two admixtures, Xypex Admix C-500 and Rheomac SF100 (silica fume), were added as an additional dry material binder in select compositions. As in the Screening Matrix, both fresh and cured properties were evaluated for the formulations. In this study, properties that were influenced by the W/DM ratio in the Screening Matrix; flow diameter, plastic viscosity, density, and compressive strength, showed consistent behavior with respect to W/DM. The leach index for highly soluble components, sodium and nitrate, were not influenced by changes in formulation or the admixtures. The leach index for both iodine and Tc-99 show an influence from the addition of the admixture, Xypex Admix C-500. Additional testing should be performed to further evaluate the influence of Xypex Admix C-500 on the leach index over a range of admixture concentrations, Cast Stone formulations, and curing and storage conditions.

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DOE OSTI2013

Breaking the Code on Challenging Waste - 13267

Mixed low-level wastes (MLLW) with no available path to treatment or disposal have been longstanding challenges for DOE facilities. Today, mixed wastes with no path to treatment or disposal frequently present themselves in the form of combinations of problematic matrixes, problematic EPA Hazardous Waste Codes, and security classification requirements. In order to successfully treat and disposition these challenging wastes, waste management personnel must be more inquisitive and challenge the status quo more than ever before. All aspects of the waste from how it was generated to how the waste is currently being managed must be revisited. Each fact, the basis of each decision, and each regulatory determination must be investigated and validated. Since many of the difficult waste streams were generated several years ago, it can be quite challenging to locate knowledgeable generators from the time of generation. Significant investigation is often required to obtain the needed information to evaluate legacy waste streams. Special attention must be paid to the little things that may not seem central to the issues being investigated. Solutions are sometimes found in these details. (authors)

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DOE OSTI2017

Progress Updates for the Hanford Waste Treatment and Immobilization Plant Project - 17275

The Waste Treatment and Immobilization Plant (WTP) Project is vital to the U.S. Department of Energy's (DOE) mission to cleanup radioactive waste at the Hanford Site, located in southeastern Washington State. The overall WTP Project objective is to design, build, and commission the facilities and systems that will treat and immobilize approximately 211 983 m{sup 3} (56 million gallons) of radioactive waste stored in Hanford's 177 underground storage tanks. The combination of technologies in the WTP makes it a first-of-a-kind facility at a scale larger than ever built for radioactive waste processing. In 2012, DOE suspended production work on the WTP's Pretreatment (PT) Facility and to a lesser degree the High-Level Waste (HLW) Facility because of unresolved technical issues associated with those facilities. DOE is proceeding with its approach to complete the WTP Project in phases, which includes the following principal actions: - Completing construction of the Low-Activity Waste (LAW) Facility, Balance of Facilities (BOF), and Analytical Laboratory (LAB), collectively known as LBL; - Implementing a direct-feed low-activity waste (DFLAW) process to treat low-activity waste independent of the PT Facility; - Completing the design and construction of the HLW Facility; - Resolving the technical issues associated with the PT Facility; - Completing all WTP facilities to enable integrated WTP operations. In March 2016, a federal judge amended the 2010 Consent Decree, establishing revised milestone dates for the Hanford WTP that generally align with DOE's plans for completing the WTP facilities in sequence. In December 2016, DOE executed a contract modification and approved a change to the project's performance baseline to complete the LBL facilities and incorporate modifications needed to accommodate direct feed of low-activity waste to the LAW Facility. The contract modification and revised performance baseline support completion of hot commissioning of the WTP facilities needed for DFLAW operations no later than December 31, 2023. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Direct Feed Options for Hanford High Level Waste Vitrification - 17318

In the high level waste (HLW) direct feed (DF) option that is under consideration for early operations of the Hanford Tank Waste Treatment and Immobilization Plant (WTP), the pretreatment facility would be bypassed in order to support an earlier start-up of the HLW vitrification facility. For HLW, this would mean that the ultrafiltration and caustic leaching operations that would otherwise have been performed in the pretreatment facility would either not be performed or would be replaced by an interim pretreatment function (in-tank washing and settling, for example). These changes could result in higher aluminum contents, higher chromium contents, a higher fraction of supernate in the HLW feed, and lower solids contents. This would likely affect glass formulations and waste loadings and have impacts on the downstream vitrification operations. In the present work, new high waste loading glass formulations have been developed and characterized for the DF HLW option. A series of scaled melter tests was then conducted to assess the impacts of increased supernate content resulting from less effective washing on wastes from candidate source tanks for the direct feed option. The melter tests were conducted at solids contents extending below the current WTP baseline, which are likely for the direct feed option. The effects on glass production rate, melter operations, and off-gas carryover were determined, as was the ability of increased bubbling to compensate for the increased evaporative load. A series of waste compositions were investigated that span the range of in-tank washing efficiencies between the baseline WTP full-wash case and the no-wash case. The tests included wastes from tank AY-102 as well as high plutonium wastes from tank A-104. The results from this work provide the initial basis for assessments of the relative merits of progressively more intensive pretreatment in HLW direct feed options versus the practical benefits of minimal pretreatment. Together, these results illustrate that the optimum washing strategy will depend on the specific tank waste composition and will need to be evaluated on a tank-by-tank basis. (authors)

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DOE OSTI2017

SRS Defense Waste Processing Facility - the Nation's Only Operating Vitrification Facility - Completes 20 Years of Radioactive Operations - 17407

Savannah River Remediation (SRR) is the current Liquid Waste contractor at the U.S. Department of Energy's (DOE) Savannah River Site (SRS), located near Aiken, South Carolina, and operates the Defense Waste Processing Facility (DWPF), along with the other SRS liquid waste facilities. The DWPF, the nation's only operating vitrification facility, completed 20 years of radioactive operations in 2016. During this period, DWPF has safely treated the high-level liquid waste at this DOE facility since radioactive operations began with the first transfer of sludge waste feed to DWPF on March 12, 1996. A little over a month later, on April 29, 1996, the first radioactive canister was poured. DWPF poured its 4,000. canister of vitrified waste on December 31, 2015. As of September 30, 2016, DWPF has removed approximately 59.9 million curies from the liquid waste system at SRS and placed them in 4,106 canisters. SRS has operationally closed 8 high-level waste tanks. The DWPF melter has poured nearly 7.14 million kg of molten glass, and the facility has used only two melters in its lifetime. Melter 1 poured 1,339 canisters (1996-2002), while melter 2 to date (September 30, 2016) has poured 2,767 canisters (2003-present). Melter 2 continues to safely and efficiently operate. In order to achieve these accomplishments, the facility has chronicled a history of continued operational improvements such as the installation of bubblers in the melter to increase production, incorporation of glass compositions to achieve higher waste loadings, and improvements to melter design to extend melter life. As SRR moves forward to the next 20 years, continued improvements as well as maintaining the aging infrastructure through a proactive system health program are key priorities. This paper chronicles 20 years of challenges, improvements and accomplishments, and discusses future operations and infrastructure upgrades including implementation of a key flowsheet change to significantly reduce hydrogen generation in the Chemical Process Cell. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Why accredit ISO/CEI 17025 a nuclear waste management laboratory? - 15112

When there is no storage, nor repository, available on a site, the entity in charge of nuclear waste management should find outlets in order to evacuate outside the site. Each outlet has his rules and waste acceptance criteria (packaging rules, physical/chemical/radiological criteria for example). Before you are able to evacuate your waste, an acceptance's process (or contract review) takes place. This process scans the entire measures implemented by the site in order to satisfy outlet's criteria. When the acceptance is obtained, shipments could be organized, and the process gets into an observation's phase. During this phase, the outlet could choose to inspect a finished nuclear waste package or to audit regularly the waste management process. In this last case, the outlet re-scans the entire documents and implemented measures. This audit lasts from one to several days. A way to tackle/approach the audit is to acknowledge, by a trusted third party (i.e. independent from the outlet), all or part of waste management system, in order to base one's reputation on this organism and win outlet's trust. For this purpose, the entity could choose certification or accreditation. Certification (ISO 9001 as often as not) aims at evaluating a system for managing the quality of one product or service. Accreditation (ISO 17025, for example) provides formal recognition for technical skills within a quality management system (clauses 5 and 4 of the ISO 17025 standard). In 2006, the LDM laboratory is the first French laboratory accredited, in accordance with ISO 17025 standard, for the characterization by gamma spectrometry on 100 L- and 200 L- radioactive waste drums (accreditation No 1-1712 - scope available on www.cofrac.fr). This accreditation allows the lab: - to update the quality management system, - to compile the entire technical file of the gamma-ray assay (description, evaluation of the uncertainties, calibration, proficiency testing programs, measurement procedures,...), - to pass previous identified constraints on to producers and nuclear waste package production, - to maintain permanent quality improvement loop, - to lighten the programs of outlets audits, through the acknowledgement of the technical competence of the laboratory to evaluate by gamma spectrometry the activity contained in our nuclear waste drums. On the other hand, the lab is re-evaluated periodically to ensure its continued compliance with requirements: within a 12 month's interval for an internal audit, with a 18 month's period for COFRAC (French sole official national accreditation body). Since 2006, the lab is accredited. His accreditation renewed in 2010. A new entire renewal audit is planned in 2015. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

The German product quality control of the compacted metallic nuclear waste - 15311

The treatment and safe disposal of spent nuclear fuel is a key challenge of nuclear energy utilization and has to be accomplished regardless the future nuclear of electricity production or phase-out policies. Until June 2005 approx. 6662 tons of spent fuel from German nuclear power plants was sent to the La Hague (France) and Sellafield (UK) plants for reprocessing. An equivalent amount of nuclear waste compounds has contractually to be returned to Germany, which is currently in progress. The returnable waste streams comprise mainly highly-active vitrified waste containing a major part of fission products and actinides, and intermediate-active super-compacted metallic waste containing reprocessed fuel element's structural materials and some technological waste from the reprocessing process itself. Shearing and dissolution of spent fuel leaves scraps of fuel metal cladding, the end-pieces and spacers which are contaminated, activated and irradiating from their life in a reactor core. In fact, this metallic intermediate level radioactive waste makes up a bit more than half of the total volume of the reprocessed waste. At La Hague the metallic waste is super-compacted. Several, typically eight, compacted discs are stacked in a stainless steel container, named CSD-C as for 'Colis Standard de Dechets Compactes'. 4104 containers of such CSD-Cs in 152 transport/storages casks have to be returned to Germany. Each one of them has to be checked and quality certified for its conformity with German national product specifications and acceptance criteria by independent national product quality control inspectors (PKS). The residue package documentation has to be complete and the characterization methods have to be comprehensive. Due to the heterogeneity of metallic waste the characterization and control of its properties is very complex. The nuclear characterization is performed by means of non-destructive measurements, where only a few key nuclides are measured directly. In fact, the proper calculation of the whole nuclear inventory relies on numerous supplementary simulations. The measurement and calculation methodology is scrutinized by PKS in detail and key points are benchmarked with our own simulations for the sake of an independent nuclear inventory declaration check. With respect to the vast amount of documentation and parameters, which have to be checked and verified, PKS had to optimize their checking procedures. For this purpose dedicated proof tools have been developed that shall be applied for selected indicative cross-checks of the relevant parameters. This contribution is dedicated to the procedure of quality control by PKS and examples of actually performed computational verifications of declared nuclear inventory. (authors)

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DOE OSTI2015

X-Ray Diffraction of Slag-Based Sodium Salt Waste Forms - 15513

Cementitious materials are used to solidify and stabilize aqueous based radioactive waste containing sodium salts. The types and proportions of cementitious ingredients used to treat aqueous radioactive waste streams containing sodium salts depend on the performance objectives for the waste forms and the compositions of the waste streams. This paper documents sample preparation and x-ray diffraction results for a series of materials made with water or highly alkaline sodium salt simulated waste water and cementitious binders. The objective of this study was to: 1) generate a base line for the evolution of the waste form as a function of time and conditions, 2) design new binders based on mineralogy of the binder, 3) understand and predict anion and cation leaching behavior of contaminants of concern, and 4) predict performance of the waste forms for which phase solubility and thermodynamic data are available. Characterization of the mineralogy is also important for understanding the buffering effects that the waste form has on infiltrating water / leachates. In summary, mixtures of Type II portland cement, Grade 100 ground granulated blast furnace slag (GGBFS) and carbon burn-out (CBO) Class F Fly ash which were hydrated with water contained hydrated phase assemblages typical of those reported in the literature. The calcium silicate hydrate phase assemblage in samples hydrated with the alkaline 4.4 M sodium salt simulated waste solution was found to be a function of the (CaO + MgO) / (SiO{sub 2} + Al{sub 2}O{sub 3}) ratio of the samples characterized. No significant differences were detected in samples cured 2 months and 14 months in sealed containers at ambient indoor temperatures. Slag and a blend of slag and cement hydrated with caustic 4.4 M Na salt solution resulted in the most crystalline matrix. In addition to poorly ordered C-S-H, these samples contained fairly well ordered C-S-H I (a precursor of 14A tobermorite) and 11 A Al-substituted tobermorite. These crystalline C-S-H phases did not form or were present in only trace amounts in slag blends containing about 45 to 62 mass percent fly ash. These slag-Class F fly ash blends had a higher silica plus alumina content relative to lime and magnesia than the blends that produced C-S-H I and Al-substituted tobermorite. The calcium silicate binder in the 10:45:45 mixture of cement: slag: fly ash was made up of poorly ordered C-S-H. The sample cured for 14 months may contain a small amount of the more crystalline calcium silicate hydrate phases. Layered double hydroxides in the hydrotalcite (magnesium-aluminum carbonate hydroxide) and hydrocalumite / AFm phases (calcium aluminum hydroxide) were present in mixtures containing slag. The specific phase(s) were not identified because these phases form solid solutions and have a considerable amount of overlap in their x-ray patterns. Sodium nitrate was the only sodium salt phase identified in x-ray diffraction patterns of the samples hydrated with salt solution. Drying during x-ray diffraction sample preparation may have resulted in precipitation of the sodium nitrate or it may have been present in the samples prior to x-ray sample preparation. Sodium sulfate, aluminate, and carbonate may have been incorporated in the structures of the layered double hydroxide (AFm) type phases. These mixed metal layered double hydroxides make up an important fraction of the matrix in the slag containing blends hydrated with caustic salt solution. They are among the few oxide-based phases that exhibit substantial, permanent anion exchange capacity [Kirkpatrick, et al. 1999, Plamer, et al., 2009, and Zhang and Reardon, 2003]. They also contribute to the structural properties of cementitious matrices [Taylor, 1997]. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Tritiated waste management opportunities based on the reduction of tritium activity and outgassing - 15607

Tritiated waste is called as such because tritium is the main radionuclide contained in this waste which is generally produced by nuclear research, fusion facilities and small producers, e.g. hospitals, research centers and industries. The management of this type of waste can be improved by using techniques that reduce the tritium activity and outgassing levels. Several options are possible depending on the waste form and the amount of tritium: temporary storage for tritium decay up to 50 years, thermal treatment such as incineration and melting, and high-integrity containers for conditioning, either considered alone or in combination with one of the other options. This paper compares a number of waste management parameters with respect to tritiated waste: the environmental impact of tritium releases, secondary waste management, volume reduction of the waste for final disposal, and the waste activity after treatment, not to mention the related safety issues, the performance of the techniques used in each option, and economic aspects. The results of this comparative work are given in this paper, with the advantages and drawbacks of each option discussed in detail. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Managing Low-Level Waste at Dounreay - 16269

Dounreay Site Restoration Limited (DSRL) is the site licence company responsible for the clean-up and demolition of Britain's former centre of fast reactor research and development. The Dounreay site requires facilities to dispose of up to 175,000 cubic metres of low level waste (LLW) that has been and will continue to be generated during the decommissioning and restoration of the site. In May 2014, construction of two new vaults for the disposal of LLW at Dounreay was completed, including the necessary roads, infrastructures and services, followed by an encapsulation plant in September 2014. The vaults are located immediately to the northeast of the Dounreay licensed site, on NDA-owned land. The facilities are for disposal of the solid LLW generated at Dounreay and the adjacent Vulcan site only. In 2004, following consultation with stakeholders and members of the public, Dounreay issued the 'Best Practicable Environmental Option (BPEO) Report' on the management of Dounreay LLW. This report assessed potential management options for LLW against a range of criteria, including technical, environmental, cost, health and safety issues. The study was conducted in line with best practice and involved consultation with a range of people, including the local community and organisations with an interest in radioactive waste management. The recommendations and conclusions from the study provided the basis for the 'Dounreay Solid LLW Overall Strategy', which was published in March 2005. A fundamental component of this strategy was the development of new below-ground disposal vaults for LLW at Dounreay. Siting the vaults on land at Dounreay avoids any need to transport the LLW away from Dounreay on public roads. Disposal at Dounreay therefore satisfies the proximity principle of managing the waste at source. Disposal at Dounreay is also consistent with UK Government Policy on LLW management, the Scottish Government's Radioactive Waste Management Policy and the Nuclear Decommissioning Authority's (NDA's) strategy on management of LLW from the UK nuclear industry. A planning application was initially lodged with the Highland Council in June 2006. The period for the determination of this application was extended to allow the Scottish Environment Protection Agency (SEPA) sufficient time to assess fully the supporting Environmental Safety Case (ESC). The planning application was re-activated in May 2008, and planning permission was granted by Highland Council in April 2009. The planning permission included 26 conditions on DSRL, covering a wide range of environmental issues such as construction noise/dust, traffic, visual impact and also a condition to develop a community fund aimed at socio-economic development of the area around the Dounreay site. After undertaking the necessary detailed design work, construction of the phase one vaults commenced in November 2011 by principal contractor, Graham Construction. Work was undertaken on both vaults simultaneously, excavating around 243,000 m{sup 3} of rock. When excavation was complete, steel reinforcement was placed and concrete poured to construct the floor slabs. This was followed by the construction of the reinforced walls and then the installation of the steel support structure and materials that form the roof. Mechanical and electrical installation then followed, in parallel with the construction of the vault apron, ramps and access roads. A pumping system and control building was also constructed to remove groundwater from around the vaults to allow the vaults to be kept dry while they are being filled with waste. A layer of excavated material, approximately 3 m thick, was placed across the existing ground surface between the vaults and the coast, to ensure the water table lies well below the new ground surface after closure of the vaults, and that any future trace contaminated groundwater flows out to the sea rather than to the soil zone. This layer is referred to in project documents as the 'enhanced geosphere'. After disposals are completed and the vaults are closed, this layer will be joined to the capping over the vaults so that the ensemble merges visually with the surrounding landscape. The original vegetation and soil mix will be returned as far as is practicable. An encapsulation plant has been constructed to fill the LLW containers with grout prior to disposal in the vault, similar to the process used at the national Low-Level Waste Repository facility (LLWR), near Drigg in Cumbria. The plant has the capacity to grout up to four containers per day. The grout fills voids in the container and helps provide chemical conditions in the vault that limit the release of radionuclides in the far future, after the vault has been penetrated by groundwater. On April 24 2015, active commissioning began at the Encapsulation Plant and LLW Vaults. Half-height ISO containers packed with low level waste are currently being filled with grout and disposed of in the LLW Vault. Also packaged Demolition Low Level Waste (DLLW) is being processed and disposed of in the Demolition Low Level Waste Vault. To date over 2500 packages of DLLW and 70 containers LLW have been disposed of. Completion of construction and the start of operating the Phase One vaults and encapsulation plant are fundamental to the successful and efficient decommissioning of the Dounreay site, upon which many millions of pounds of investment and hundreds of jobs rely.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Cement Solidification of Residual Metal Oxide from Radioactive Waste

A lot of uranium contaminated soil and concrete waste is generated from the dismantlement of uranium conversion facilities. There are several disposal methods for radioactive materials such as regulation exemption, decontamination and long-term storage. It is necessary for us to achieve the permanent disposal of radioactive waste. The methods for solidifying radioactive waste include cement solidification, asphalt solidification, and polymer solidification. Cement solidification has several advantages such as well-known materials and technologies, various applications, and reasonable cost. On the contrary, it has certain disadvantages such as a large volume expansion of solidification for radioactive waste. Because cement solidification has excellent structural strength and a shielding effect, it is widely used for the permanent disposal of radioactive waste. Cement solidification operates as a barrier against the leaching of nuclides during radioactive waste disposal. To evaluate the criteria of KORAD (Korea Radioactive Waste Agency), integrity tests on the solidification of radioactive waste need to be performed. In this study, to inspect the criteria of KORAD, we carried out the cement solidification of residual metal oxide from radioactive waste. An integrity inspection such as the pressure strength, inundation, and leaching test on cement solidification was performed. We also analyzed the experimental results according to the criteria of KORAD. The calculation of the leachability index on the primary and secondary cement solidification for leaching detection of nuclides is carried out. The integrity of primary and secondary cement solidification is evaluated. (authors)

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DOE OSTI2019

Concept for Operation of the Belgian Geological Disposal Facility for Category B and C Wastes - 19247

For more than 40 years, ONDRAF/NIRAS (the federal agency responsible for managing radioactive waste and enriched fissile materials in Belgium) has been studying the concept of a geological disposal facility (GDF) in poorly indurated clays as an option for the long-term management of both category B waste (low-level and intermediate-level long-lived waste - LILW-LL) and category C waste (high-level waste - HLW and spent fuel). In the absence of a policy decision, the proposed Belgian reference option for disposal of such waste is a GDF in poorly indurated clays, located at a depth of between 200 m and 600 m. Within the framework of the first Safety and Feasibility Case (SFC-1) undertaken by ONDRAF/NIRAS, the organization is updating its former GDF layout, which dates back from 2003. The SFC-1 GDF concept is characterized by a central zone in which shafts, joined by a central connecting gallery, provide two surface connections. On opposing sides of the central gallery, four (two on each side) access galleries branch off to connect the disposal areas for B- and C-waste, respectively. Disposal dead end galleries, 400 m long, branch off perpendicular to the access galleries. Both waste areas will be operated during two separate time frames. The current underground layout has been significantly modified from the previous 2003 layout. The modifications are largely based on several R and D studies with regard to the future operations of the GDF carried out by BGE Technology GmbH in support to ONDRAF/NIRAS. This paper focuses on major operational processes such as waste transport and emplacement. In addition, the current conceptual design of the backfilling and closure concept is presented. Optimization studies on the further development of the operational process are also presented. (authors)

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DOE OSTI2020

Development of the 3 m{sup 3} Low Level Radioactive Waste Container in Taiwan - 20037

Chinshan Nuclear Power Plant stopped its business operation in July, 2019, and Kuosheng Nuclear Power Plant will also be permanently shut down in 2023. Nuclear Reactor Facilities Regulation Act requires a nuclear facility must be dismantled in 25 years after being permanently ceased operation. According to the approved decommissioning plan of Chinshan Nuclear Power Plant and the proposed one of Kuosheng Nuclear Power Plant, the 3 m{sup 3} low level radioactive waste container is planned to store the low level radioactive wastes with higher radiation produced during the decommissioning. However, 3 m{sup 3} low level radioactive waste containers are not used or developed in Taiwan before. To meet such need, Institute of Nuclear Energy Research develops a new type of 3 m{sup 3} low level radioactive waste container and performs the corresponding analyses and tests based on Guidelines of Applying the Usage License of Low Level Radioactive Waste Containers and Regulations for the Safe Transport of Radioactive Material. In this article, it provides the information of the low level radioactive waste containers that have been approved or is under reviewing in Taiwan, and so do the design specifications of the developed 3 m{sup 3} waste container. Besides, the numerical evaluation and test results of the developed 3 m{sup 3} waste container are also included in this article. Based on the evaluation and test results, the developed 3 m{sup 3} waste container satisfies the requirements of the storage container and industrial package type 2. Institute of Nuclear Energy Research will submit the usage license applications to the authority this year, and the usage licenses would be granted in 2021, expectedly. (authors)

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