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Feedback from International Decommissioning Planning Projects - 20269

Decommissioning of Nuclear Power Plants is an increasing reality in many countries with nuclear power. Already many years before final shutdown, planning of the decommissioning work is usually started in order to understand for example the costs, waste volume capacity and licensing work needed. In many countries it is required to have a final decommissioning plan before any decommissioning work is allowed to start. Westinghouse has been involved in international decommissioning planning for more than 30 years, where the first 20 years was mostly funding and preliminary plans due to that decommissioning was many years away. However, as many more plants are now moving into decommissioning, the focus has shifted to final decommissioning plans with budgets instead of funding plans, strategic decisions instead of choices, waste volume specifications instead of estimates and a defined approach instead of what ifs. Westinghouse has developed a methodology based on the planning and execution of many decommissioning projects in Europe, America and Asia. That international experience and understanding helps the customers to adopt a systematic planning process where strategy development is done with the end state in mind, where the goals are clearly outlined and the starting point described. The process and steps to reach the desired end state need to be clearly defined and all options/alternatives evaluated from a cost/benefit point of view. This is followed by a thorough and quantitative risk analysis in order to give solid recommendations and credible conclusions. One important feedback has been that real decommissioning experience from reference projects is invaluable for an efficient and timely progress in order not to spend unnecessary time and effort on understanding the barriers and links between different activities in decommissioning. Skilled project managers or engineers can drive the project forward, but without decommissioning experience previously gained, valuable lessons learned are not used and the same mistakes can be made again. Many utilities have invested in knowledge transfer from other decommissioning projects through site visits, exchange programs and supplier meetings, which is a valuable asset and way of implementing lessons learned from performed decommissioning projects into the planning process. (authors)

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Low Level Radon Concentration and Radon Daughter Dose Monitoring in an Arid-zone Waste Store - 20081

Radioactive waste containing predominantly soil and rubble from the clean-up of a former uranium and thorium minerals processing research facility is currently stored at the Woomera Test Range in South Australia awaiting final disposal. Prior to commencing any manipulation of the nearly 10,000 waste drums (205 L or 55 gallon each), radon concentration monitoring and radon daughter dosimetry has been undertaken to (i) gain understanding of the temporal and spatial variation of radon and its daughters within the storage facility, which could provide insights into the radiological source term and (ii) confirm that radiological risks to workers are below the guideline values. A time series of nearly 12 months of detailed observations shows low level radon concentrations with annual average values that are slightly above the average radon concentrations in Australian homes. Peak values in summer are about three times higher than the annual average based on the CRM measurements. Both radon concentrations and annual effective dose are well below the regulatory limits. (authors)

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Leaching of Au, Ag, and Pd from waste printed circuit boards of mobile phone by iodide lixiviant after supercritical water pre-treatment

Highlights: • We report a novel process for recovering Au, Ag, and Pd from waste PCBs. • The effect of SCWO on the leaching of Au, Ag, and Pd in waste PCBs was studied. • SCWO was highly efficient for enhancing the leaching of Au, Ag, and Pd. • The optimum leaching parameters for Au, Ag, and Pd in iodine–iodide were studied. - Abstract: Precious metals are the most attractive resources in waste printed circuit boards (PCBs) of mobile phones. In this work, an alternative process for recovering Au, Ag, and Pd from waste PCBs of mobile phones by supercritical water oxidation (SCWO) pre-treatment combined with iodine–iodide leaching process was developed. In the process, the waste PCBs of mobile phones were pre-treated in supercritical water, then a diluted hydrochloric acid leaching (HL) process was used to recovery the Cu, whose leaching efficiency was approximately 100%, finally the resulting residue was subjected to the iodine–iodide leaching process for recovering the Au, Ag, and Pd. Experimental results indicated that SCWO pre-treatment temperature, time, and pressure had significant influence on the Au, Ag, and Pd leaching from (SCWO + HL)-treated waste PCBs. The optimal SCWO pre-treatment conditions were 420 °C and 60 min for Au and Pd, and 410 °C and 30 min for Ag. The optimum dissolution parameters for Au, Pd, and Ag in (SCWO + HL)-treated PCBs with iodine–iodide system were leaching time of 120 min (90 min for Ag), iodine/iodide mole ratio of 1:5 (1:6 for Ag), solid-to-liquid ratio (S/L) of 1:10 g/mL (1:8 g/mL for Ag), and pH of 9, respectively. It is believed that the process developed in this study is environment friendly for the recovery of Au, Ag, and Pd from waste PCBs of mobile phones by SCWO pre-treatment combined with iodine–iodide leaching process.

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Assessment of Options for the Treatment of Nitrate Salt Wastes at Los Alamos National Laboratory

This paper summarizes the methodology used to evaluate options for treatment of the remediated nitrate salt waste containers at Los Alamos National Laboratory. The method selected must enable treatment of the waste drums, which consist of a mixture of complex nitrate salts (oxidizer) improperly mixed with sWheat Scoop®1, an organic kitty litter and absorbent (fuel), in a manner that renders the waste safe, meets the specifications of waste acceptance criteria, and is suitable for transport and final disposal in the Waste Isolation Pilot Plant located in Carlsbad, New Mexico. A Core Remediation Team was responsible for comprehensively reviewing the options, ensuring a robust, defensible treatment recommendation. The evaluation process consisted of two steps. First, a prescreening process was conducted to cull the list on the basis for a decision of feasibility of certain potential options with respect to the criteria. Then, the remaining potential options were evaluated and ranked against each of the criteria in a consistent methodology. Numerical scores were established by consensus of the review team. Finally, recommendations were developed based on current information and understanding of the scientific, technical, and regulatory situation. A discussion of the preferred options and documentation of the process used to reach the recommended treatment options are presented.

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Waste Management 2016 300-296 Soil Remediation Project. Developing Innovative Solutions to Overcome Design Challenges - 16229

The 324 Building is a non-reactor, Hazard Category 2, Nuclear Facility located in the 300 Area of the U.S. Department of Energy (DOE) Hanford Site, located near Richland, Washington. The building is undergoing deactivation, decommissioning, decontamination, and demolition by Washington Closure Hanford (WCH) under the River Corridor Closure Contract. During the Cold War Era, the facility was used for chemical and radionuclide processing associated with nuclear weapons production. Historical records indicate that in October 1986, approximately 516 L of a concentrated liquid waste stream containing cesium-137 and strontium-90 was spilled onto the floor of B-Cell. B-Cell is located in the Radiochemical Engineering Complex (REC) of the 324 Building. It is estimated that this spill contained approximately 1.3 million curies of radioactivity. Some fraction of the spill was recovered; however, an unknown quantity was released into the soil beneath the 324 Building through an undetected breach in the B-Cell floor sump. Unknown quantities of water were used after the spill to wash items contained in B-Cell. When B-Cell was taken out of service, wastes being removed from B-Cell were placed in Grout Containers (GCs) and stabilized using a particulate grout. In the course of grouting activities, grout was spilled on the floor of B-Cell, eventually filling the sump. The primary scope of this project is to design, test, and implement a process capable of removing and isolating contaminated soil below B-Cell. This remediation effort has been divided into two removal zones identified as the primary and secondary removal zones. Based on current assumptions and estimates, the primary phase of this cleanup effort will remove roughly 156.45 m{sup 3} of highly contaminated materials. B-Cell debris, particulate grout, and the floor slab and liner comprise about 21.24 m{sup 3} of the total. The remaining 135.21 m3 will be contaminated soil. All of this material must be disposed of within the existing 324 Facility or packaged for disposal at the Hanford site. The objective of this project is to remove the B-Cell floor and highly contaminated underlying soils from beneath B-Cell using remote equipment to reduce the radiation levels and allow open-air excavation for the remaining contaminated soil. The challenges and subsequent technological solutions encountered from the Conceptual Design phase to the 100% Design Issued For Fabrication/Issued For Construction (IFF/IFC) consisted in part of the following: Design an efficient remote cleanup strategy for highly contaminated Low Level Waste (LLW) given the unknowns. Maintain contamination control down to 18 feet below B-Cell in a highly radioactive environment (12,870 R/Hr Cs-137 and Sr-90). Perform facility upgrades to enable remediation process implementation while maintaining integrity of pre-existing facility structure and license. Designing first-of-a-kind remote equipment for use in high radiation areas. Developing first-of-a-kind operations sequences for remote equipment while minimizing usage of original facility equipment (i.e. cranes, manipulators, etc.). Conducting time-motion studies of operations sequences to maximize efficiency while keeping dose As Low As Reasonably Achievable (ALARA) in a high dose environment. Construction of a full scale mockup to conduct performance testing in a clean environment and train the work force to operate the equipment. Overcoming scope, budget and schedule constraints associated with a fast-track DOE project. (authors)

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Development of characterization protocol for mixed liquid radioactive waste classification

Mixed liquid organic waste generated from health-care and research activities containing tritium, carbon-14, and other radionuclides posed specific challenges in its management. Often, these wastes become legacy waste in many nuclear facilities and being considered as ‘problematic’ waste. One of the most important recommendations made by IAEA is to perform multistage processes aiming at declassification of the waste. At this moment, approximately 3000 bottles of mixed liquid waste, with estimated volume of 6000 litres are currently stored at the National Radioactive Waste Management Centre, Malaysia and some have been stored for more than 25 years. The aim of this study is to develop a characterization protocol towards reclassification of these wastes. The characterization protocol entails waste identification, waste screening and segregation, and analytical radionuclides profiling using various analytical procedures including gross alpha/ gross beta, gamma spectrometry, and LSC method. The results obtained from the characterization protocol are used to establish criteria for speedy classification of the waste.

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Downgrading Nuclear Facilities to Radiological Facilities

Based on inventory reductions and the use of alternate storage facilities, the Sandia National Laboratories (SNL) downgraded 4 SNL Hazard Category 3 (HC-3) nuclear facilities to less-than-HC-3 radiological facilities. SNL’s Waste Management and Pollution Prevention Department (WMPPD) managed the HC-3 nuclear facilities and implemented the downgrade. This paper will examine the downgrade process,

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Soap production: A green prospective

Highlights: • Reducing the amount of different types of municipal waste. •Recycling waste to obtain a useful green soap, which is good quality and cheap. • Inspiring community projects meant to convey recycling and ecological approach. • Contributing to families’ economy and stimulating the use of healthier products. - Abstract: A green prospective based on the reuse of waste materials such as almond shells, orange peel and used cooking oil to manufacture soap is presented. In Portugal, thousands of tons of waste are generated from used cooking oil and production of nut shells’ residues is growing every year. In addition, the high consumption of citrus fruits, oranges in particular, generates large amounts of citrus peel. Therefore, it is necessary to diversify reuse mechanisms of these wastes, in order to make them back into raw materials. Complying with this trend, this work was carried out by processing and grinding almond shells, treating used oil, processing orange peels and extracting limonene, formulating and producing soap, and performing an acceptance study of the final product. Results validated a high potential of the idea in the field of environmental education, so it can be replicated in practical classes. It can also be useful for waste management, and it can support the development of community projects on an ecological approach.

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Analysis of DWPF Sludge Batch 6 (Macrobatch 7): Pour Stream Glass Samples

The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 6 (SB6), also referred to as Macrobatch 7 (MB7), in June 2010. SB6 is a blend of the heel of Tank 40 from Sludge Batch 5 (SB5), H-Canyon Np transfers and SB6 that was transferred to Tank 40 from Tank 51.1 SB6 was processed using Frit 418. Sludge is received into the DWPF Chemical Processing Cell (CPC) and is processed through the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator Tank (SME). The treated sludge slurry is then transferred to the Melter Feed Tank (MFT) and fed to the melter. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program (GPCP) and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. The DWPF requested various analyses of radioactive glass samples obtained from the melter pour stream during processing of SB6 as well as reduction/oxidation (REDOX) analysis of MFT samples to determine the impact of Argon bubbling. Sample analysis followed the Task Technical and Quality Assurance Plan (TTQAP) and an Analytical Study Plan (ASP). Four Pour Stream (PS) glass samples and two MFT slurry samples were delivered to the Savannah River National Laboratory (SRNL) from the DWPF. Table 1-1 lists the sample information for each pour stream glass sample. SB6 PS3 (S03472) was selected as the official pour stream sample for SB6 and full analysis was requested. This report details the visual observations of the as-received SB6 PS No.3 glass sample as well as results for the chemical composition, Product Consistency Test (PCT), radionuclide content, noble metals, and glass density. REDOX results will be provided for all four pour stream samples and vitrified samples of MFT-558 and MFT-568A. Where appropriate, data from other pour stream samples will be provided.

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Modeling Crystalline Silicotitanate Performance to Support TCCR Batch 1 and Batch 1A Operations (U)

The objective of this work is to calculate the expected ion-exchange (IX) column performances of the Tank Closure Cesium Removal (TCCR) system for Tank 10H closure. Savannah River National Laboratory (SRNL) utilized: (1) the ZAM computer program (Version 4) developed by Texas A&M University for prediction of cesium loading on the crystalline silicotitanate (CST) media, (2) the OLI Studio™ software (Version 9.6) from OLI Systems to estimate waste solution properties, and (3) the VERSE-LC code (Version 7.8) developed by Purdue University to calculate column performances. The study specifically evaluates the TCCR system performance to process Tank 10H Batch 1A supernate.

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Cast Stone Formulation for Nuclear Waste Immobilization at Higher Sodium Concentrations

Low activity radioactive waste at U.S. Department of Energy sites can be immobilized for permanent disposal using cementitious waste forms. This study evaluated waste forms produced with simulated wastes at concentrations up to twice that of currently operating processes. The simulated materials were evaluated for their fresh properties, which determine processability, and cured properties, which determine waste form performance. The results show potential for greatly reducing the volume of material. Fresh properties were sufficient to allow for processing via current practices. Cured properties such as compressive strength meet disposal requirements. Leachability indices provide an indication of expected long-term performance.

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Vapor-Liquid Partitioning of Methylmercury Compounds: Fundamental Data to Support the Savannah River Site Liquid Waste System: Henry's Law, Solubility and Vapor Pressure Determination for Representative Methylmercury Compounds

The Savannah River Site (SRS) Liquid Waste System (LWS) contains approximately 66 tons of mercury within the liquids, salts, and sludges that are currently being processed into final wasteforms for disposal. Mercury concentrations within the system exceed those typically experienced in environmental or industrial systems; thus, management of mercury compounds continues to be a priority for SRS. In the LWS, waste is vitrified into a borosilicate glass wasteform that contains most of the radioactivity, while the lower activity solutions are dispositioned in a low-level grout wasteform, or “saltstone”. The alkaline, high ionic strength caustic wastes are pumped, evaporated, and otherwise managed throughout the LWS and Defense Waste Processing Facility (DWPF) as they are stored and prepared for conversion to the final wasteforms. Because of the complexity of this system, a key component of effective mercury management in the LWS requires analysis of mercury in various physical phases. The high concentration of mercury within the SRS LWS has the potential to generate vapor-phase contamination. Elemental mercury (Hg 0 ), dimethylmercury ((CH 3 ) 2 Hg), and methylmercury (CH 3 Hg + ) are among species known or suspected to contribute to the flux of mercury from liquid to vapor phase (Iverfeldt and Lindquist, 1982). Chemical speciation affects not only mercury behavior in LWS operations but may also affect the performance of mercury treatment and removal technologies in the LWS.

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Can a radiological event impact a criticality safety program: Lessons learned at WIPP - 14350

Any accident at a nuclear facility impacts all departments involved in the operation of the facility. The radiological release that occurred on February 14, 2014 has halted all waste disposal operations at the Waste Isolation Pilot Plant (WIPP) and has resulted in detailed reviews of all safety management programs. This paper describes the initial reactions to the event, criticality issues with recovery operations, corrective actions for the criticality safety program from independent reviews, and the importance of configuration management to criticality safety. The nature, location, and timing of the event resulted in significant delays in determining the cause of the event. During this time multiple theories on the initiator of the event were postulated including and inadvertent criticality event. However, the timing of the event, average fissile gram equivalent (FGE) loading of waste containers compared to analyzed FGE loads, and isotopic analysis of fixed air samples taken during the event proved it was not an inadvertent criticality event. Recovery activities such as replacing contaminated high efficiency particulate (HEPA), filters and disposing of site generated waste have also been reviewed to prevent criticality safety issues. Post event corrective actions in all departments have impacted how future operations at WIPP will be conducted, including corrective actions for the criticality safety program. The event has also proven the importance of configuration management and the effect a change can have on a process. (authors)

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Canonsburg, Pennsylvania, UMTRCA Title I Disposal Site 35 Years Postclosure (1985-2020) - 20288

The DOE Office of Legacy Management (LM) Canonsburg, Pennsylvania, Disposal Site is a former uranium ore-processing site located in the borough of Canonsburg and an important part of US history. For 46 years (1911-1957), the site processed ore, for vanadium, radium and then uranium. Soils and groundwater beneath the site were contaminated by the milling operations. The 15-hectare (37-acre) site is owned by the US government and managed by LM, which serves as the federal land manager and steward of cultural, historical, and natural resources at Cold War legacy sites that have been successfully cleaned up to remedial standards. In 1978 Congress directed DOE to remediate 22 inactive uranium-ore processing sites under the Uranium Mill Tailings Radiation Control Act (UMTRCA), in accordance with standards promulgated by the US Environmental Protection Agency in Title 40 CFR Part 192 (40 (CFR 192). Subpart B of 40 CFR 192 regulates the cleanup of contaminated groundwater at the processing sites. The radioactive materials were encapsulated in US Nuclear Regulatory Commission (NRC)-approved disposal cells, and the NRC general license for UMTRCA Title I sites is established in 10 CFR 40.27. Canonsburg was the first site to be remediated under UMTRCA. From 1983 to 1985, the Canonsburg site underwent cleanup under this US law that provides for the safe and environmentally sound disposal, long-term stabilization, and control of uranium mill tailings in a manner that minimizes or eliminates radiation health hazards to the public. The site is unique in that it is one of only two UMTRCA Title I sites in the eastern half of the United States, the other site being the Burrell, Pennsylvania, Title I Disposal Site, a vicinity property to the Canonsburg site. It has been 35 years since the disposal cell at the Canonsburg site was completed. This paper discusses the many efforts that have been taken to maintain the site and presents some of the lessons learned concerning the long-term care of the site. The discussion includes but is not limited to the following: (1) Preserving site history so the community can better appreciate and understand the site. (2) Community outreach efforts, and the importance of maintaining transparency of environmental monitoring results with the community. (3) Pursuing reuse opportunities that offer the potential for improving the community. (4) Remaining engaged with the local community. (5) Remaining proactive in the long-term care of the site. (authors)

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TRIPLICATE SODIUM IODIDE GAMMA RAY MONITORS FOR THE SMALL COLUMN ION EXCHANGE PROGRAM

This technical report contains recommendations from the Analytical Development (AD) organization of the Savannah River National Laboratory (SRNL) for a system of triplicate Sodium Iodide (NaI) detectors to be used to monitor Cesium-137 ({sup 137}Cs) content of the Decontaminated Salt Solution (DSS) output of the Small Column Ion Exchange (SCIX) process. These detectors need to be gain stabilized with respect to temperature shifts since they will be installed on top of Tank 41 at the Savannah River Site (SRS). This will be accomplished using NaI crystals doped with the alpha-emitting isotope, Americium-241({sup 241}Am). Two energy regions of the detector output will be monitored using single-channel analyzers (SCAs), the {sup 137}Cs full-energy {gamma}-ray peak and the {sup 241}Am alpha peak. The count rate in the gamma peak region will be proportional to the {sup 137}Cs content in the DSS output. The constant rate of alpha decay in the NaI crystal will be monitored and used as feedback to adjust the high voltage supply to the detector in response to temperature variation. An analysis of theoretical {sup 137}Cs breakthrough curves was used to estimate the gamma activity expected in the DSS output during a single iteration of the process. Count rates arising from the DSS and background sources were predicted using Microshield modeling software. The current plan for shielding the detectors within an enclosure with four-inch thick steel walls should allow the detectors to operate with the sensitivity required to perform these measurements. Calibration, testing, and maintenance requirements for the detector system are outlined as well. The purpose of SCIX is to remove and concentrate high-level radioisotopes from SRS salt waste resulting in two waste streams. The concentrated high-level waste containing {sup 137}Cs will be sent to the Defense Waste Processing Facility (DWPF) for vitrification and the low-level DSS will be sent to the Saltstone Production Facility (SPF) to be incorporated into grout.

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Investigation of Reactive Transport and Coupled THMC Processes in the EBS

Geological repositories for disposal of high-level nuclear waste generally rely on a multibarrier system to isolate radioactive waste from the biosphere. The multi-barrier system typically consists of the natural barrier system (NBS), which includes the repository host rock and its surrounding subsurface environment, and the engineered barrier system (EBS). The EBS represents the man-made, engineered materials placed within a repository, including the waste form, waste canisters, buffer materials, backfill, and seals (OECD, 2003). The EBS plays a significant role in the containment and long-term retardation of radionuclide release.

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Tin(II)apatite: Synthesis, Characterization, and Challenge with Pertechnetate - 16194

We will report on the synthesis of stannous apatite and the subsequent reductive capacity attained, using a cerium (IV) titration method. Also reported will be the distribution coefficients derived using a Hanford tank supernatant simulant of 7.8 M sodium with chrome (VI) as a contaminant of concern, as a potential non-isotopic oxy-anion and technetium surrogate for reductive capacity determinations. Pertechnetate will be subjected to reduction using hydroxylamine, glucose and threose. The purpose is to take the first step in laboratory analyses to determine the amount of non-pertechnetate species in Hanford tank waste. A proposed sequestration mechanism will be presented as to why the reacted tin(II)apatite does not release technetium once reacted. SEM and EDS spectra will be presented on the crystalline apatite. Areas of incorporation of the tin(II)apatite will be presented for the Hanford tank waste process. (authors)

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Multiple stakeholders in multi-criteria decision-making in the context of Municipal Solid Waste Management: A review

Highlights: • We review Municipal Solid Waste Management studies with focus on multiple stakeholders. • We focus on studies with multi-criteria decision analysis methods and discover their trends. • Most studies do not offer solutions for situations where stakeholders compete for more benefits or have unequal voting powers. • Governments and experts are the most participated stakeholders and AHP is the most dominant method. - Abstract: Municipal Solid Waste Management (MSWM) is a complicated process that involves multiple environmental and socio-economic criteria. Decision-makers look for decision support frameworks that can guide in defining alternatives, relevant criteria and their weights, and finding a suitable solution. In addition, decision-making in MSWM problems such as finding proper waste treatment locations or strategies often requires multiple stakeholders such as government, municipalities, industries, experts, and/or general public to get involved. Multi-criteria Decision Analysis (MCDA) is the most popular framework employed in previous studies on MSWM; MCDA methods help multiple stakeholders evaluate the often conflicting criteria, communicate their different preferences, and rank or prioritize MSWM strategies to finally agree on some elements of these strategies and make an applicable decision. This paper reviews and brings together research on the application of MCDA for solving MSWM problems with more focus on the studies that have considered multiple stakeholders and offers solutions for such problems. Results of this study show that AHP is the most common approach in consideration of multiple stakeholders and experts and governments/municipalities are the most common participants in these studies.

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