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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

International Approaches for Nuclear Waste Disposal in Geological Formations: Report on Fifth Worldwide Review

An important issue for present and future generations is the final disposal of spent nuclear fuel. Over the past over forty years, the development of technologies to isolate both spent nuclear fuel (SNF) and other high-level nuclear waste (HLW) generated at nuclear power plants and from production of defense materials, and low- and intermediate-level nuclear waste (LILW) in underground rock and sediments has been found to be a challenging undertaking. Finding an appropriate solution for the disposal of nuclear waste is an important issue for protection of the environment and public health, and it is a prerequisite for the future of nuclear power. The purpose of a deep geological repository for nuclear waste is to provide to future generations, protection against any harmful release of radioactive material, even after the memory of the repository may have been lost, and regardless of the technical knowledge of future generations. The results of a wide variety of investigations on the development of technology for radioactive waste isolation from 19 countries were published in the First Worldwide Review in 1991 (Witherspoon, 1991). The results of investigations from 26 countries were published in the Second Worldwide Review in 1996 (Witherspoon, 1996). The results from 32 countries were summarized in the Third Worldwide Review in 2001 (Witherspoon and Bodvarsson, 2001). The last compilation had results from 24 countries assembled in the Fourth Worldwide Review (WWR) on radioactive waste isolation (Witherspoon and Bodvarsson, 2006). Since publication of the last report in 2006, radioactive waste disposal approaches have continued to evolve, and there have been major developments in a number of national geological disposal programs. Significant experience has been obtained both in preparing and reviewing cases for the operational and long-term safety of proposed and operating repositories. Disposal of radioactive waste is a complex issue, not only because of the nature of the waste, but also because of the detailed regulatory structure for dealing with radioactive waste, the variety of stakeholders involved, and (in some cases) the number of regulatory entities involved.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Advanced Experimental Technique for Radiation Damage Effects in Nuclear Waste Forms: Neutron Total Scattering Analysis

For the past 30 years, the development of durable materials for radionuclide immobilization has been driven by efforts to dispose of wastes generated by the nuclear fuel cycle [National Research Council, ‘Waste Forms Technology and Performance: Final Report’, the National Academies Press, Washington D.C., 2011]. Many materials have been developed, but there still exist large gaps in the knowledge of fundamental modes of waste form degradation in repository environments. An important aspect of waste form science is the behavior of the materials under intense irradiation from decaying actinides and fission products. This irradiation induces a wide range of defects and disorder, the details of which depend on the specific waste form material. At the present time, it is not fully explained how radiation effects will influence the performance of nuclear waste forms and their long-term retention of fission products and actinides under operational conditions. The complex defect behavior and radiation damage must be understood over a range of length scales, from the initial atomic-scale defect structure to the long-range observable material modification. This is particularly challenging and requires advanced characterization techniques. This contribution describes how pair distribution function (PDF) analysis obtained from neutron total scattering experiments can be applied in the research field of waste form science to uniquely characterize radiation effects in a wide range of materials, including crystalline complex oxides and waste glasses. Neutron scattering strength does not have an explicit Z-dependence; this allows access to many low-Z elements, such as oxygen, that cannot be accurately studied with X-rays. In many cases, this can permit a detailed analysis of both cation (often high-Z) and anion (often low-Z) defect behavior. In contrast to traditional crystallography, which relies on long-range order, PDF analysis probes the local defect structure, including changes in site occupation, coordination, and bond distance. This is particularly important when characterizing aperiodic waste glasses with no long-range order at all. In contrast to X-ray characterization which requires very little sample mass (~0.1 mg), neutron characterization (even at state-of-the-art spallation facilities) requires relatively large sample mass (~50 - 100 mg). In conclusion, obtaining this quantity is challenging for studies of irradiated materials, but by tailoring our experimental approach to use high-energy ions (GeV) with very high penetration depth, we are able to produce the required mass.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Broad Scope Analysis of the MOX Fuel Fabrication Facility and Disposition of Plutonium

The Plutonium Management and Disposition Agreement (PMDA), signed into effect in 2000, between the United States and Russia, was made in a post-Cold War effort to dispose of military-useable material, a combined 68 metric tons of weapon-grade plutonium. The United States has been pursuing a disposition path through conversion to MOX fuel for use in current commercial nuclear reactors. The Savannah River Site’s (SRS) MOX Fuel Fabrication Facility (MFFF, the MOX facility) began construction in 2007 with intent to uphold the United States’ half of the agreement. In the FY2017 budget, the Obama administration proposed a termination of the MFFF after lack of sufficient funding to the project, which caused paralyzing delays and drastically inflated costs for completion. Prolonged construction, piled costs, and divergent policies between shifting administrations have resulted in turmoil between the United States and Russia, culminating with the proposed abandonment of the MOX disposition path and being used as a justification by Russia to withdraw from the agreement until the United States is able to identify a clear plan, fulfilling the requirements set forth by the PMDA. The leading expedient to replace MOX fuel is the dilute and dispose method, which leaves uncertainty for the SRS MFFF and places reliance on New Mexico’s Waste Isolation Pilot Plant (WIPP) to legally accommodate the additional and unaccounted waste. If the current capacity of WIPP proves insufficient, construction of a new repository for transuranic waste would only incur additional costs and regulatory burdens, along with extreme schedule delays. Future proposed adaptations of the SRS MFFF include pit production, which would require complex and costly alterations to be operational at the current site. Restrictions from Congress, such as mandating the proposed solution to abandoning MFFF must equal half of the costs required for completing the facility, have fueled severely underestimated quotations of these alternatives. The increasing pressure from both Russia and South Carolina government officials to make a decision could result in the pursuit of another doomed project, leaving the United States with even further sunk costs and strained international relations. This paper will summarize the involved history of the MFFF, explicate the recourses to salvage the facility for pit production, and provide postulation on the dilute and dispose method.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Investigation of Coupled Processes and Impact of High Temperature Limits in Argillite Rock: FY17 Progress. Predecisional Draft

The focus of research within the Spent Fuel and Waste Science and Technology (SFWST) (formerly called Used Fuel Disposal) Campaign is on repository-induced interactions that may affect the key safety characteristics of EBS bentonite and an argillaceous rock. These include thermal-hydrologicalmechanical- chemical (THMC) process interactions that occur as a result of repository construction and waste emplacement. Some of the key questions addressed in this report include the development of fracturing in the excavation damaged zone (EDZ) and THMC effects on the near-field argillaceous rock and buffer materials and petrophysical characteristics, particularly the impacts of temperature rise caused by waste heat. This report documents the following research activities. Section 2 presents THM model developments and validation, including modeling of underground heater experiments at Mont Terri and Bure underground research laboratories (URLs). The heater experiments modeled are the Mont Terri FE (Full-scale Emplacement) Experiment, conducted as part of the Mont Terri Project, and the TED in heater test conducted in Callovo-Oxfordian claystone (COx) at the Meuse/Haute-Marne (MHM) underground research laboratory in France. The modeling of the TED heater test is one of the Tasks of the DEvelopment of COupled Models and their VAlidation against EXperiments (DECOVALEX)-2019 project. Section 3 presents the development and application of thermal-hydrological-mechanical-chemical (THMC) modeling to evaluate EBS bentonite and argillite rock responses under different temperatures (100 °C and 200 °C). Model results are presented to help to understand the impact of high temperatures on the properties and behavior of bentonite and argillite rock. Eventually the process model will support a robust GDSA model for repository performance assessments. Section 4 presents coupled THMC modeling for an in situ test conducted at Grimsel underground laboratory in Switzerland in the Full-Scale Engineered Barrier Experiment Dismantling Project (FEBEX-DP). The data collected in the test after almost two decades of heating and two dismantling events provide a unique opportunity of validating coupled THMC models and enhancing our understanding of coupled THMC process in EBS bentonite. Section 5 presents a planned large in-situ test, “HotBENT,” at Grimsel Test Site, Switzerland. In this test, bentonite backfilled EBS in granite will be heated up to 200 °C, where the most relevant features of future emplacement conditions can be adequately reproduced. Lawrence Berkeley National Laboratory (LBNL) has very actively participated in the project since the very beginning and have conducted scoping calculations in FY17 to facilitate the final design of the experiment. Section 6 presents present LBNL’s activities for modeling gas migration in clay related to Task A of the international DECOVALEX-2019 project. This is an international collaborative activity in which DOE and LBNL gain access to unique laboratory and field data of gas migration that are studied with numerical modeling to better understand the processes, to improve numerical models that could eventually be applied in the performance assessment for nuclear waste disposal in clay host rocks and bentonite backfill. Section 7 summarizes the main research accomplishments for FY17 and proposes future work activities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Fault Stability Perturbation by Thermal Pressurization and Stress Transfer Around a Deep Geological Repository in a Clay Formation

The increase of temperature of a low-permeability, fluid-saturated media may trigger significant thermal pressurization, where the expansion of pore fluid cannot be accommodated by the thermal expansion of the pore space. With the aid of a coupled thermohydromechanical numerical simulator, we investigate the possible impact of thermal pressurization during the life of a deep geological repository (DGR) for high-level radioactive waste, characterized by the emplacement of radioactive material-filled canisters in a series of parallel tunnels, excavated in a low-permeability clay formation. We represent the fault as a planar structure embedded in a thermoporoelastic material and shear activation evaluated by a strain-softening Mohr-Coulomb failure criterion. The results show that stress changes caused by temperature and thermal pressurization of a rock mass around the emplacement tunnels may trigger a slip event on a fault plane in proximity of the geological disposal site: Rupture nucleates at depth, hundreds of meters below the DGR. Stress transfer plays a key role, while a direct hydraulic connection between the repository and the nucleation zone is not necessary in order to trigger rupture. A low stress ratio may favor the occurrence of slip up to a distance of 600 m of the fault from the outermost tunnel. Furthermore, these results highlight the need of investigating hydromechanical properties and local stress conditions at depth to characterize the geomechanical response of weak planes located in the surroundings of a high-level radioactive waste repository and to provide sufficient knowledge for the safe development of the DGR site.

58 GEOSCIENCES↗
DOE OSTI2020

Quest for Continuous Design and Safety Optimization: A Look at Cires, the French Very Low Level Waste Disposal - 20010

Cires, the French repository for very low-level waste (VLLW) managed by Andra, has a fifteen-year history of enhancing design and disposal operations, responding to producers' needs by optimizing existing facilities and providing new facilities as needed. Its flexibility allows the facility to adapt to changes in waste delivery flows, to dispose of larger waste volumes without using more space, to integrate new concepts and even envisage disposal capacity extension. The article traces the ever evolving history of technical and operational optimization of this unique disposal facility situated in Eastern France. (authors)

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DOE OSTI2015

Portland cement paste behaviour under gamma radiation field in geological repository - 15272

The Radioactive Waste Management Laboratory at the Nuclear and Energy Research Institute in Sao Paulo, Brazil (IPEN-CNEN-SP), is developing the concept of a repository for disposition of disused sealed radioactive sources in a deep borehole in which cement paste is intended to be used as a backfill between the steel casing and the geological formation around the borehole. The cement paste is intended to function as structural material, a blockage against the transport of water between the different strata of the geological setting, and as an additional barrier against the migration of radionuclides. The assessment of the cement paste behaviour in long term is necessary to increase the confidence that the material will perform as required during the service life of the facility. Under radiation, pore water present in cement paste suffers radiolysis and some products can be highly reactive, as electrons and hydroxyl radicals and hydrogen peroxide. These radiolysis products will interact with cement paste and its hydration products, forming a wide range of compounds. Previous work focused on assessing the synergic effects of radiation and environmental conditions, like higher temperature and aggressive groundwater chemicals, on cement paste. Examination of samples detected small differences in mineralogy and microstructure. However, the radiation dose to which samples were exposed (400 kGy) was only a fraction of the expected accumulated dose in cement paste until complete decay of some sealed sources in the repository. Therefore, several thousand kilo-grays must be tested the possibility of radiation to alter cement mineralogy and, as a consequence, its durability The objective of this research is to evaluate the changes in mineralogy and microstructure of irradiated cement paste specimens, in an attempt to establish a relationship between irradiation and durability. Cubic Portland cement specimens were casted and exposed to the gamma radiation of a 60 Co multipurpose irradiator, at IPEN-CNEN/SP. The effects of doses between 1000 kGy and 4000 kGy are evaluated by X-Ray Diffraction, Scanning Electron Microscopy, X-ray micro tomography, neutrongraphy and neutron scattering. This report presents the changes in cement paste mineralogy when samples are examined by XRD. Results indicate that durability can be affected because of the observed changes. Results are expected to indicate the induced changes in cement paste under high radiation doses and to provide hints on its behaviour in the long-term and service-life. The results can be used in safety analysis for deep geological repositories that use cement based materials as engineered barriers. (authors)

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DOE OSTI2017

Waste Minimization of HyBRID Decontamination Process - 17060

The Korea Atomic Energy Research Institute (KAERI) developed a Hydrazine base reductive metal ion decontamination (HyBRID) and related pre-oxidation techniques for use in decommissioning a nuclear facility. Inorganic chemicals were used to increase the integrity of the radioactive waste in a repository. Decomposition and precipitation of residual reagents with inorganic chemicals were introduced to minimize the final waste volume. The waste volume was calculated theoretically and demonstrated in a small-scale waste treatment system and bench-scale candle filtration system. The total waste volume of the HyBRID process was less than 1/10 of that of commercial processes, and spent ion-exchange resins were significantly reduced to 5% of that of commercial processes. (authors)

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

Advanced Fuel Cycle Cost Basis – 2017 Edition

This report, commissioned by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the DOE Nuclear Technology Research and Development (NTRD) Program (previously the Fuel Cycle Research and Development (FCRD) and the Advanced Fuel Cycle Initiative (AFCI)). The report describes the NTRD cost basis development process, reference information on NTRD cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for numerous fuel cycle cost modules (modules A-O) as well as cost modules for a number of reactor types (R modules). The fuel cycle cost modules were developed in the areas of natural uranium mining and milling, thorium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, managed decay storage, recycled product storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste. Since its inception, this report has been periodically updated. The last such internal document was published in August 2015 while the last external edition was published in December of 2009 as INL/EXT-07-12107 and is available on the Web at URL: www.inl.gov/technicalpublications/Documents/4536700.pdf. This current report (Sept 2017) is planned to be reviewed for external release, at which time it will replace the 2009 report as an external publication. This information is used in the ongoing evaluation of nuclear fuel cycles by the NE NTRD program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Enthalpies of formation of polyhalite: A mineral relevant to salt repository

Polyhalite is an important coexisting mineral with halite in salt repositories for nuclear waste disposal, such as Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. The thermal stability of this mineral is a key knowledge in evaluating the integrity of a salt repository in the long term, as water may release due to thermal decomposition of polyhalite. Previous studies on structural evolution of polyhalite at elevated temperatures laid the basis for detailed calorimetric measurements. Using high-temperature oxide-melt drop-solution calorimetry at 975 K with sodium molybdate as the solvent, we have determined the standard enthalpies of formation from constituent sulfates (ΔH° f,sul ), oxides (ΔH° f,ox ) and elements (ΔH° f,ele ) of a polyhalite sample with the composition of K 2 Ca 2 Mg(SO 4 ) 4 ·1.95H 2 O from the Salado formation at the WIPP site. The obtained results are: ΔH° f,sul = -152.5 ± 5.3 kJ/mol, ΔH° f,ox = -1926.1 ± 10.5 kJ/mol, and ΔH° f,ele = -6301.2 ± 9.9 kJ/mol. Furthermore, based on the estimated formation entropies of polyhalite, its standard Gibbs free energy of formation has been derived to be in the range of -5715.3 ± 9.9 kJ/mol to -5739.3 ± 9.9 kJ/mol. In conclusion, these determined thermodynamic properties provide fundamental parameters for modeling the stability behavior of polyhalite in salt repositories.

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DOE OSTI2018

Buffer Material Degradation in High Level Radioactive Waste Disposal - 18454

The concept of geological disposal for high-level radioactive waste (HLW) is based on a multi-barrier system comprising the natural geological barrier that repository host rock provides and its surroundings, as well as an engineered barrier system (EBS) that includes waste form, waste canisters, buffer materials, and backfill. The multi-barrier system is expected to perform its desired functions to isolate waste from the biosphere and achieve HLW disposal. Because of the interaction processes of radionuclide decay heat with hydraulic and water-mineral reactions in the radioactive waste repository, the bentonite buffer degradation resulting from the external temperature and pressure warrants investigation. However, seldom addressed in Taiwan is the influence of thermo-hydro-chemical (T-H-C) processes on buffer material degradation in the EBS of HLW disposal repositories as related to smectite clay dehydration. Therefore, we adopted the chemical kinetic model of smectite dehydration to calculate the amount of water expelled from smectite clay minerals caused by higher temperatures of waste decay heat. We discovered that smectite clay could cause volume shrinkage because of a loss of interlayer water, which can lead to bentonite buffer compression. The results of this study may be used in advanced research on the evolution of bentonite degradation for both performance assessments and safety analyses of final HLW disposal in Taiwan. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Autonomous Inspection of Geologic Repositories: Current State of the Art and Future Directions

Geological repositories for nuclear waste, including spent nuclear fuel, present a significant challenge for traditional International Atomic Energy Agency (IAEA) safeguards tools due to their inaccessibility and demanding operational conditions. The IAEA has been working closely with Member State organizations currently involved in repository construction and planning, including Euratom, the Finnish and Swedish regulatory authorities, and relevant facility operators. However, the verification challenge remains unsolved, and there persists an out- standing need for tools and approaches that will help the IAEA verify that no nuclear material is diverted from a repository environment. The challenge is also not static. Activities must encompass verification of the design, prior to and during the construction/operation phase, and post backfill. Throughout these various phases, it is imperative that the IAEA maintains a continuity of knowledge (CoK) of all material including information on material inventory and flow. This paper highlights these challenges and outlines how they might be addressed by using remote or autonomous vehicles. Specifically, we discuss the current state of the art in robotic autonomy for known or partially known environment mapping and patrolling, as well as shared autonomy, where humans collaborate with closed loop autonomation to complete tasks. We explore the feasibility of using rovers for these verification tasks, along with the challenges associated with system implementation. Hardware and software suggestions are provided based on the adoption of similar technologies in other comparable areas and ability to close technical gaps. Lastly, human-robotic interactions are considered based on the challenges of the environment of the repository and effective deployment and continued operation of the robot system.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Review of the Safety Analyses of the Konrad Repository - 20530

The Bundesgesellschaft fuer Endlagerung (BGE) is the license holder of all geological repository sites in Germany since 2017. Besides the construction of the Konrad repository for the emplacement of low- and intermediate-level waste (LILW), the operation of the Asse II mine including the retrieval of the emplaced waste, and the preparation for the closure of the Morsleben repository, BGE is responsible to carry out the operative tasks of selecting a site for a high-level waste (HLW) repository. The Konrad repository is Germany's first repository for radioactive waste with negligible heat generation, which was fully licensed under atomic law within the scope of a plan approval procedure from the original application in 1982 until the license was granted in 2002 and finally confirmed by the Federal Administrative Court in 2007. The emplacement of LILW was licensed up to a volume of 303,000 m{sup 3} with a total β/γ-activity of 5 x 10{sup 18} Bq and a total α-activity of 1.5 x 10{sup 17} Bq. Currently, the former iron ore mine Konrad is being converted to a repository and is expected to come into operation in 2027. But what if there is no safety case in place nor a regulatory framework tailored for a change of documents in order to ensure the safety according to the state-of-the-art of science and technology during the evolution of a long-term project such as a geological repository? In this presentation, a review procedure of the safety analyses, that have been the basis of the plan approval decision for the Konrad repository at that time, will be described. Furthermore, the first results of this graduated review procedure for updating the safety analyses will be presented. (authors)

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

A National Disposal Facility for Radioactive Waste for Iraq - 16610

Radioactive waste resulted in the past in Iraq from a number of nuclear activities and facilities all of them seriously damaged or destroyed during the Gulf war. The task of decommissioning such facilities and managing the resulting waste is being supported by the International Atomic Energy Agency (IAEA), the European Union (EU), and the United States (US). A national policy for radioactive waste management for Iraq has been drafted by the Ministry of Science and Technology (MoST); a draft strategy for radioactive waste management implementation is in preparation. In this framework, Iraq is considering locating a radioactive waste disposal facility at the Al-Tuwaitha site, near the capital city of Bagdad. In the framework of its Instrument for Nuclear Safety Cooperation the EU has recently started a cooperation project named 'Feasibility Study, Basic Design, and Engineering Design of an Engineered Radioactive Waste Disposal Facility' for Iraq (Project IQ4.01/11). The project aims at producing the design of a repository for the radioactive waste resulting from the decommissioning activities and from the continuing use of radioisotopes in medicine and the oil and gas industry in the country. The contract for this cooperation project was awarded by the European Commission in April 2015 to the consortium of two international engineering companies, Nukem Technologies Engineering Services GmbH and DBE Technology GmbH, both of Germany, and Javys a.s., the national waste management agency of Slovakia. The planned Iraqi near-surface repository is constrained by the complex security conditions in the country. This has led to the decision of using whenever possible the Al-Tuwaitha site, were most of the waste is located, as the repository location. The project work will therefore focus on assessing whether and which part of the waste expected to result from the decommissioning and remediation activities is amenable for surface disposal in an appropriately engineered facility. In addition, careful review of the site characteristics shall provide enough elements to justify and validate the political decision of disposing of the waste in Al Tuwaitha. Thereafter, the project work will focus on the conceptual and later the engineering design of the repository. Finally, a Safety Analysis will provide the basis for regulatory decisions about the further steps for realization of the Iraqi final disposal facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Scenario development for safety assessment in deep geologic disposal of high-level radioactive waste and spent nuclear fuel: A review

Radiation and radioactive substances result in the production of radioactive wastes which require safe management and disposal to avoid risks to human health and the environment. To ensure permanent safe disposal, the performance of a deep geological repository for radioactive waste is assessed against internationally agreed risk-based standards. Assessing postclosure safety of the future system's evolution includes screening of features, events, and processes (FEPs) relevant to the situation, their subsequent development into scenarios, and finally the development and execution of safety assessment (SA) models. Global FEP catalogs describe important natural and man-made repository system features and identify events and processes that may affect these features into the future. By combining FEPs, many of which are uncertain, different possible future system evolution scenarios are derived. Repository licensing should consider both the reference or “base” evolution as well as alternative futures that may lead to radiation release, pollution, or exposures. Scenarios are used to derive and consider both base and alternative evolutions, often through production of scenario-specific SA models and the recombination of their results into an assessment of the risk of harm. Furthermore, while the FEP-based scenario development process outlined here has evolved somewhat since its development in the 1980s, the fundamental ideas remain unchanged. A spectrum of common approaches is given here (e.g., bottom–up vs. top–down scenario development, probabilistic vs. bounding handling of uncertainty), related to how individual numerical models for possible futures are converted into a determination as to whether the system is safe (i.e., how aleatoric uncertainty and scenarios are integrated through bounding or Monte Carlo approaches).

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