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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Refinement of Modeling Techniques for the Structural Evaluation of Hanford Single-Shell Nuclear Waste Storage Tanks

ABSTRACT Several tanks at the Hanford Site (in Washington State, USA) belong to the first generation of underground nuclear waste storage tanks known as single shell tanks (SSTs). These tanks were constructed between 1943 and 1964 and are well beyond their design life. This article discusses the structural analysis approach and modeling challenges encountered during the ongoing analysis of record (AOR) for evaluating the structural integrity of the SSTs. There are several geometrical and material nonlinearities and uncertainties to be dealt with while performing the modern finite element analysis of these tanks. The analysis takes into account the temperature history of the tanks and allowable mechanical operating loads of these tanks for proper estimation of creep strains and thermal degradation of material properties. The loads prescribed in the AOR models also include anticipated loads that these tanks may see during waste retrieval and closure. Due to uncertainty in a number of inputs to the models, sensitivity studies were conducted to address questions related to the boundary conditions to realistically or conservatively represent the influence of surrounding tanks in a tank farm, the influence of backfill excavation slope, the extent of backfill and the total extent of undisturbed soil surrounding the backfill. Because of the limited availability of data on the thermal and operating history for many of the individual tanks, some of the data was assumed or interpolated. However, the models developed for the analysis of record represent the bounding scenarios and include the loading conditions that the tanks were subjected to or anticipated. The modeling refinement techniques followed in the AOR resulted in conservative estimates for force and moment demands at various sections in the concrete tanks. This article discusses the modeling aspects related to Type-II and Type-III SSTs. The modeling techniques, methodology and evaluation criteria developed for evaluating the structural integrity of SSTs at Hanford are in general applicable to any similar tanks or underground concrete storage structures.

Waste Managament, Single Shell Tanks, Structural E↗
NASA NTRS1985

Thermal sensing for characterizing the contents of waste storage drums

The present investigation is concerned with the feasibility to employ remote sensing for the characterization of the contents of liquid chemical waste storage drums. Philipson et al. (1981) had found that, when the air temperature is changing rapidly, differences in the thermal inertia of the drum contents can lead to detectable differences in the skin temperature of the drums. Thus, postsunset, airborne thermal remote sensing could potentially provide some level of discrimination among chemical storage drums. Discrimination should be possible among steel drums filled largely with: (1) aqueous solvents, (2) organic solvents, or (3) clay packing materials. The response of a drum filled with clay packing materials should be similar to that of an empty drum. The reported study had the objective to verify the theoretical findings, taking into account the use of a hand-held infrared radiometer. It was found that under the proper conditions the temperature differences among drums with the three different types of contents will be significant and consistent.

Philpot, W. D.↗
DOE OSTI2020

Default Radioactive and Hazardous Waste Management (RHWM) Dose Conversion Factors (DCFs) and Weighting Factors and plutonium-239 equivalent Curie (PE-Ci) methodology for acceptance into the Waste Storage Facilities

DOE-STD-5506-2007 (Ref. 1) is the standard that provides the methodology for preparing hazard analysis and accident analysis for transuranic (TRU) waste facilities. The Waste Storage Facilities hazard and accident analysis for the DSA is developed in accordance with DOE-STD- 5506-2007. Two related issues need to be addressed when implementing DOE-STD-5506-2007 regarding the calculation of consequences from TRU waste releases in accident analysis: Consequences determined in accident analysis must be modeled using inhalation Dose Conversion Factors (DCFs) that are consistent with those from the International Commission on Radiological Protection Publication 72 (ICRP-72, Ref. 2).; The use of a statistical Material at Risk (MAR) representing the distribution of radiological loading of TRU waste containers based on the use of a plutonium-239 equivalent Curie (PE-Ci) methodology. The PE-Ci methodology simplifies accident analysis by allowing one to model releases based on the Curie value of one standard nuclide (e.g., Pu-239) instead of needing to model releases for every potential nuclide of interest. This is accomplished by normalizing the Curie value of all nuclides to a single Curie value of a standard nuclide by using a Weighting Factor that is a ratio of the DCFs associated with the nuclides and that of the standard nuclide. By using the ICRP-72 DCFs to calculate PE-Ci values used in the statistical MAR, the accident analysis that uses that statistical MAR will be consistent with the consequence methodology identified in DOE-STD- 5506-2007.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Application of QMU to the design of a nuclear waste storage tank

Defining the reliability of complex physical systems is crucial for system certification and guaranteeing safety, but is often complicated by the presence of many uncertainties and system sub-components. One such example is the certification of the Hanford Waste Treatment Plant, where radioactive and toxic chemical waste that spontaneously generates flammable gases is stored in large tanks. We introduce the QMU methodology to quantify the system safety of waste storage tanks in the event of a loss of tank ventilation, which poses a safety risk to workers. We also show how QMU can be used as a design tool to find operating regimes that maintain a desired confidence level. A simplified physical model is used to analyze the build-up of hydrogen gas in storage tanks and conduct an uncertainty propagation study to quantify the operating margin with respect to the flammability conditions. A demonstration problem is shown in which the radiation loading in a waste tank is iteratively modified until a desired confidence ratio is achieved. The results show that the resulting designs are sensitive to the use of interval or full QMU to achieve a confidence ratio, and it is not always possible to realize a design at a particular confidence ratio.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Verification Survey Report of the South Waste Tank Farm Training/Test Tower and hazardous Waste Storage Lockers at the West Valley Demonstration Project, West Valley, New York (Final Report)

A team from ORAU's Independent Environmental Assessment and Verification Program performed verification survey activities on the South Test Tower and four Hazardous Waste Storage Lockers. Scan data collected by ORAU determined that both the alpha and alpha-plus-beta activity was representative of radiological background conditions. The count rate distribution showed no outliers that would be indicative of alpha or alpha-plus-beta count rates in excess of background. It is the opinion of ORAU that independent verification data collected support the site's conclusions that the South Tower and Lockers sufficiently meet the site criteria for release to recycle and reuse.

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

Use of Automation in Low Level Waste Storage and Management - 19227

The Australian Nuclear Science and Technology Organisation (ANSTO) generates Low Level Waste (LLW) and Intermediate Level Waste (ILW) from various site operations such as the operation of a research Reactor and the manufacture of Molybdenum 99. The Low Level Solid Waste (LLSW) generated is stored at ANSTO before and after conditioning. The current process flow involves the storage of LLSW waste before conditioning on racks, handled using forklifts. When the unconditioned waste is ready for conditioning, it is transferred to another facility, where volume reduction and packaging is carried out. Once conditioned, the waste is placed in further storage, pending dispatch to the proposed Australian National Radioactive Waste Management Facility (NRWMF). A new purpose built facility has been constructed, adjacent to the LLSW processing facility, for the storage and management of LLSW in both unconditioned and conditioned forms. This new facility will streamline the waste management process and significantly enhance the safety and efficiency of the process. In line with international best practises, the new facility utilises automation solutions to minimise the build footprint, and minimise the use of forklifts and other mobile plant. Most importantly, it applies the principle of ALARA (As Low as Reasonably Achievable) to minimise operator activities in areas where radioactive material is stored and eliminates double handling of the waste. The automation for unconditioned waste utilises an Automated Storage and Retrieval System (ASRS), a vertical stacker crane within high density racking, to provide easy storage and retrieval for processing and/or routine inspections. This system allows operators to control it from outside of the storage area whilst utilising integrated software that allows for the required tracking of the unconditioned waste. From the ASRS storage area, unconditioned waste will have a smooth transition to the processing facility which is located adjacent to the ASRS system. Once conditioned, the waste is catalogued and returned to a designated storage area within the new facility using an automated building gantry crane with a target delivery system to eliminate the need for operators in the main storage bay. The waste will be held here in preparation for dispatch to the future repository. The utilisation of the ASRS for the purpose of storing LLSW is an innovative use of automation that will eliminate operators from active areas, reduce dose rates received and integrate with the waste tracking software. Furthermore, the system has significant advantages with operational efficiencies increased, reduced double handling of nuclear waste and maximum use of building footprint. (authors)

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

Testing of Novel Application of GOTHIC{sup TM} to Modelling of Hydrogen Distribution in Intermediate Level Waste Storage Facilities - 16499

One of the strategies for treatment of intermediate level waste in the United Kingdom is to retrieve the waste from the legacy storage area, condition in a suitable storage container and then hold on-site in an intermediate level waste store until permanent storage in a geological disposal facility becomes available. During storage, hydrogen may be generated inside the containers at a slow rate by a number of mechanisms. The containers can be equipped with vents that release hydrogen and other gasses into the intermediate level waste store, a building which would be constructed to high standards to minimise ingress of moisture. With the low number of air changes associated with the building design there is the potential for gas to accumulate over time. There is, therefore, a requirement to be able to predict future gas concentrations inside the building over a range of operating conditions to ensure that a safe operating envelope is maintained. The usual approach to this problem is to model the evolution of the hydrogen distribution by use of computational fluid dynamics tools. In a situation where weeks of real time need to be simulated in order to understand the slow build-up of a very small volumetric concentration of hydrogen, the use of computational fluid dynamics is not practicable within the time scales demanded by industry. This paper investigates a new approach to tackling this problem using the thermal hydraulic code GOTHIC{sup TM} (GOTHIC{sup TM} incorporates technology developed for the electric power industry under the sponsorship of EPRI, the Electric Power Research Institute). (authors)

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

Current status and performance assessment for the Techa cascade of reservoirs - liquid radioactive waste storage facility

The Techa cascade of water reservoirs is the world's largest open storage facility for liquid low-level radioactive waste. Its capacity is about 360 mln. m{sup 3}, it occupies an area of more than 65 km{sup 2}, the total activity accumulated in the water and sediments is about 6.10{sup 15} Bq. The major challenge facing the Techa cascade is virtually uncontrollable water level changes. The water level rise can cause significant pollution of the environment. From the late 1990's onwards, the issue of the Techa cascade safety assurance is considered to be one of the major challenges pertaining to nuclear legacy for 'Mayak' and Russia as a whole. Unlike other industrial water reservoirs the Techa cascade liquidation is estimated as highly unrealistic. The main objectives of the paper are: - brief results summary of the practical works on safety improvement at the Techa cascade carried out over the past decade; - introduction the works on the Techa cascade performance assessment; - determination of the existing risks and strategic areas for solving the problem. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Aluminum phosphate ceramics for waste storage

The present disclosure describes solid waste forms and methods of processing waste. In one particular implementation, the invention provides a method of processing waste that may be particularly suitable for processing hazardous waste. In this method, a waste component is combined with an aluminum oxide and an acidic phosphate component in a slurry. A molar ratio of aluminum to phosphorus in the slurry is greater than one. Water in the slurry may be evaporated while mixing the slurry at a temperature of about 140-200.degree. C. The mixed slurry may be allowed to cure into a solid waste form. This solid waste form includes an anhydrous aluminum phosphate with at least a residual portion of the waste component bound therein.

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

A Versatile Remediation Module for Remote Repair of Spent Nuclear Fuel and High-Level Waste Storage Containers

Oak Ridge National Laboratory (ORNL) successfully demonstrated the Versatile Remediation Module (VRM), a prototype module designed and built by ORNL for on-site remote repair of welded stainless steel storage containers for spent nuclear fuel and high-level radioactive waste. This paper describes the VRM prototype and its design features and components to support continued long-term storage or off-site transportation of spent nuclear fuel and high-level radioactive waste currently stored in storage containers. A remote (100 ft away from the simulated radiative environment) demonstration of the VRM was successfully performed on a full-scale mock-up welded stainless steel canister. The VRM is designed with features to accommodate remediation techniques beyond those currently selected and described in this paper. Therefore, many of the VRM’s features may benefit other remote nuclear or nonnuclear applications. The VRM is envisioned to serve as a development center to facilitate and enhance further development of new remediation technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

This comprehensive report provides definitive volume, mass, and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. Calcine composition data are required for regulatory compliance (such as permitting and waste disposal), future treatment of the caline, and shipping the calcine to an off-Site-facility (such as a geologic repository). This report also contains a description of the calcine storage bins. The Calcined Solids Storage Facilities (CSSFs) were designed by different architectural engineering firms and built at different times. Each CSSF has a unique design, reflecting varying design criteria and lessons learned from historical CSSF operation. The varying CSSF design will affect future calcine retrieval processes and equipment. Revision 4 of this report presents refinements and enhancements of calculations concerning the composition, volume, mass, chemical content, and radioactivity of calcined waste produced and stored within the CSSFs. The historical calcine samples are insufficient in number and scope of analysis to fully characterize the entire inventory of calcine in the CSSFs. Sample data exist for all the liquid wastes that were calcined. This report provides calcine composition data based on liquid waste sample analyses, volume of liquid waste calcined, calciner operating data, and CSSF operating data using several large Microsoft Excel (Microsoft 2003) databases and spreadsheets that are collectively called the Historical Processing Model. The calcine composition determined by this method compares favorably with historical calcine sample data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Corrosion Control Measures For Liquid Radioactive Waste Storage Tanks At The Savannah River Site

The Savannah River Site has stored radioactive wastes in large, underground, carbon steel tanks for approximately 60 years. An assessment of potential degradation mechanisms determined that the tanks may be vulnerable to nitrate- induced pitting corrosion and stress corrosion cracking. Controls on the solution chemistry and temperature of the wastes are in place to mitigate these mechanisms. These controls are based upon a series of experiments performed using simulated solutions on materials used for construction of the tanks. The technical bases and evolution of these controls is presented in this paper.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

The Performance of Underground Radioactive Waste Storage Tanks at the Savannah River Site: A 60-Year Historical Perspective

The Savannah River Site produced weapons-grade materials for nearly 35 years between 1953 and 1988. The legacy of this production is nearly 37 million gallons of radioactive waste. Since the 1950s, the liquid waste has been stored in large, underground carbon steel waste tanks. During the past 20 years, the site has begun to process the waste so that it may be stored in vitrified and grout forms, which are more suitable for long-term storage. Over the history of the site, some tanks have experienced leakage of the waste to the secondary containment. This article is a review of the instances of leakage and corrosion degradation that the tanks and associated equipment have experienced since the first tanks were built. Furthermore, the activities that the site has taken to mitigate the degradation and manage the service life of the tank for its anticipated lifetime are reviewed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

This comprehensive report provides definitive information regarding the volume, mass, and composition (chemical and radioactivity) of the calcined waste stored at the Idaho Nuclear Technology and Engineering Center on the Idaho National Laboratory Site. The calcine composition data are required for current regulatory compliance and safe operation of the INTEC’s six Calcined Solids Storage Facilities (CSSFs), as well as future retrieval, treatment, and shipping the calcine to an off-Site facility (such as a geologic repository). This report also contains a description of the calcine storage bins that compose each of the CSSFs, emphasizing the features that may be of concern to future calcine retrieval. The CSSFs were designed by different architectural engineering firms and built at different times. Each CSSF has a unique design, reflecting varying design criteria and lessons learned from historical CSSF operation. The varying CSSF designs will affect future calcine retrieval processes and equipment. This revision (5) of the report presents refinements of the databases used to calculate the volume, mass, chemical content, and radioactivity of calcined waste produced and stored within the CSSFs. This revision corrects minor errors and inconsistencies noted in an extensive validation and verification effort of the prior version of the Microsoft Excel workbooks upon which the calcine composition described in this report is based. Historical calcine samples are insufficient in number and scope of analysis to fully characterize the entire inventory of calcine in the CSSFs. Sample data exist for all of the liquid wastes that were calcined. This report provides calcine composition data based on liquid waste sample analyses, the volume of liquid waste calcined, calciner operating data, and CSSF operating data using several large Microsoft Excel (Microsoft 2010) databases and spreadsheets. The Excel spreadsheets are collectively called the Historical Processing Model (HPM). The calcine composition determined by this method compares favorably with historical calcine sample data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

WASTE ISOLATION PILOT PLANT (WIPP): THE NATIONS' SOLUTION TO NUCLEAR WASTE STORAGE AND DISPOSAL ISSUES

In the southeastern portion of my home state of New Mexico lies the Chihuahauan desert, where a transuranic (TRU), underground disposal site known as the Waste Isolation Pilot Plant (WIPP) occupies 16 square miles. Full operation status began in March 1999, the year I graduated from Los Alamos High School, in Los Alamos, NM, the birthplace of the atomic bomb and one of the nation’s main TRU waste generator sites. During the time of its development and until recently, I did not have a full grasp on the role Los Alamos was playing in regards to WIPP. WIPP is used to store and dispose of TRU waste that has been generated since the 1940s because of nuclear weapons research and testing operations that have occurred in Los Alamos, NM and at other sites throughout the United States (U.S.). TRU waste consists of items that are contaminated with artificial, man-made radioactive elements that have atomic numbers greater than uranium, or are trans-uranic, on the periodic table of elements and it has longevity characteristics that may be hazardous to human health and the environment. Therefore, WIPP has underground rooms that have been carved out of 2,000 square foot thick salt formations approximately 2,150 feet underground so that the TRU waste can be isolated and disposed of. WIPP has operated safely and successfully until this year, when two unrelated events occurred in February 2014. With these events, the safety precautions and measures that have been operating at WIPP for the last 15 years are being revised and improved to ensure that other such events do not occur again.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Refinement of Modeling Techniques for the Structural Evaluation of Hanford Single-Shell Nuclear Waste Storage Tanks - 12288

The single-shell tanks at the Hanford Site (in Washington State, USA) were constructed between 1943 and 1964 and are well beyond their estimated 25 year design life. This article discusses the structural analysis approach and modeling challenges encountered during the ongoing analysis of record for evaluating the structural integrity of the single-shell tanks. There are several geometrical and material nonlinearities and uncertainties to be dealt with while performing the modern finite element analysis of these tanks. The analysis takes into account the temperature history of the tanks and allowable mechanical operating loads for proper estimation of creep strains and thermal degradation of material properties. The loads prescribed in the analysis of record models also include anticipated loads that may occur during waste retrieval and closure. Due to uncertainty in a number of modeling details, sensitivity studies were conducted to address questions related to boundary conditions that realistically or conservatively represent the influence of surrounding tanks in a tank farm, the influence of backfill excavation slope, the extent of backfill and the total extent of undisturbed soil surrounding the backfill. Because of the limited availability of data on the thermal and operating history for many of the individual tanks, some of the data was assumed or interpolated. However, the models developed for the analysis of record represent the bounding scenarios and include the loading conditions that the tanks were subjected to or anticipated. The modeling refinement techniques followed in the analysis of record resulted in conservative estimates for force and moment demands at various sections in the concrete tanks. This article discusses the modeling aspects related to Type-II and Type-III single-shell tanks. The modeling techniques, methodology and evaluation criteria developed for evaluating the structural integrity of single-shell tanks at Hanford are in general applicable to other similar tanks or underground concrete storage structures. This article presented the details of the finite element models and analysis approach followed during the ongoing effort to establish structural integrity of single shell tanks at the Hanford site. The details of the material constitutive models applicable to the underground Hanford concrete tanks that capture the thermal and creep induce degradation are also presented. The thermal profiles were developed based on the available tank temperature data for the Type II and Type III single-shell tanks, and they were chosen to yield conservative demands under the thermal and operating loads analysis of these tanks. Sensitivity studies were conducted to address two issues regarding the soils modeled around the single-shell tanks. The results indicate that excluding the boundary separating the backfill soil from the undisturbed soil will result in conservative demands (plots 14b and 14c green lines for circumferential Demand/Capacity ratios). The radial extent study indicated that the soil model extending to 240 ft gave more conservative results than the model with 62 ft of soil (plots 17a and 17c magenta lines for hoop Demand/Capacity ratios). Based on these results, a 240 ft far-field soil boundary with backfill throughout the lateral extent was recommended and used for the finite element models used in the Type-II and Type-III analyses of record. The modeling effort and sensitivity studies discussed in this article helped in developing bounding models for the structural integrity evaluation of single shell tanks at the Hanford site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗