HYDRAULIC AND PHYSICAL PROPERTIES OF SALTSTONE GROUTS AND VAULT CONCRETES

International Nuclear Information System (INIS)

Dixon, K.; Harbour, J.; Phifer, M.

2008-01-01

The Saltstone Disposal Facility (SDF), located in the Z-Area of the Savannah River Site (SRS), is used for the disposal of low-level radioactive salt solution. The SDF currently contains two vaults: Vault 1 (6 cells) and Vault 4 (12 cells). Additional disposal cells are currently in the design phase. The individual cells of the saltstone facility are filled with saltstone. Saltstone is produced by mixing the low-level radioactive salt solution, with blast furnace slag, fly ash, and cement (dry premix) to form a dense, micro-porous, monolithic, low-level radioactive waste form. The saltstone is pumped into the disposal cells where it subsequently solidifies. Significant effort has been undertaken to accurately model the movement of water and contaminants through the facility. Key to this effort is an accurate understanding of the hydraulic and physical properties of the solidified saltstone. To date, limited testing has been conducted to characterize the saltstone. The primary focus of this task was to estimate the hydraulic and physical properties of three types of saltstone and two vault concretes. The saltstone formulations included saltstone premix batched with (1) Deliquification, Dissolution, and Adjustment (DDA) salt simulant (w/pm 0.60), (2) Actinide Removal Process (ARP)/Modular Caustic Side Solvent Extraction Unit (MCU) salt simulant (w/pm 0.60), and (3) Salt Waste Processing Facility (SWPF) salt simulant (w/pm 0.60). The vault concrete formulations tested included the Vault 1/4 concrete and two variations of the Vault 2 concrete (Mix 1 and Mix 2). Wet properties measured for the saltstone formulations included yield stress, plastic viscosity, wet unit weight, bleed water volume, gel time, set time, and heat of hydration. Hydraulic and physical properties measured on the cured saltstone and concrete samples included saturated hydraulic conductivity, moisture retention, compressive strength, porosity, particle density, and dry bulk density. These properties

REDUCTION CAPACITY OF SALTSTONE AND SALTSTONE COMPONENTS

Energy Technology Data Exchange (ETDEWEB)

Roberts, K.; Kaplan, D.

2009-11-30

The duration that saltstone retains its ability to immobilize some key radionuclides, such as technetium (Tc), plutonium (Pu), and neptunium (Np), depends on its capacity to maintain a low redox status (or low oxidation state). The reduction capacity is a measure of the mass of reductants present in the saltstone; the reductants are the active ingredients that immobilize Tc, Pu, and Np. Once reductants are exhausted, the saltstone loses its ability to immobilize these radionuclides. The reduction capacity values reported here are based on the Ce(IV)/Fe(II) system. The Portland cement (198 {micro}eq/g) and especially the fly ash (299 {micro}eq/g) had a measurable amount of reduction capacity, but the blast furnace slag (820 {micro}eq/g) not surprisingly accounted for most of the reduction capacity. The blast furnace slag contains ferrous iron and sulfides which are strong reducing and precipitating species for a large number of solids. Three saltstone samples containing 45% slag or one sample containing 90% slag had essentially the same reduction capacity as pure slag. There appears to be some critical concentration between 10% and 45% slag in the Saltstone formulation that is needed to create the maximum reduction capacity. Values from this work supported those previously reported, namely that the reduction capacity of SRS saltstone is about 820 {micro}eq/g; this value is recommended for estimating the longevity that the Saltstone Disposal Facility will retain its ability to immobilize radionuclides.

Leaching of saltstones containing fly ash

International Nuclear Information System (INIS)

Barnes, M.W.; Roy, D.M.; Langton, C.A.

1985-01-01

Two types of fly ash were incorporated in saltstones designed for potential encapsulation of Savannah River Plant low level defense waste. These fly ashes have some cementitious properties while at the same time their presence in substitution for cement slows early hydration. Class C fly ash has a high calcium content and is considered cementitious; Class F fly ash has a low calcium content and is not classified as cementitious. Leach tests were performed and physical properties were measured for saltstones containing each class, to see the differences in the effect of the fly ashes. The four waste ions nitrate, nitrite, sodium and sulfate were shown to leach by diffusion. Effective diffusivities were determined for these ions. Data for nitrate, the most important species from the environmental point of view, are shown in Table A. Saltstones made with Class C fly ash have substantially lower leach rates than those made with Class F fly ash. The leach rates, and therefore the square roots of the effective diffusivities, have been found to be proportional to the pore surface area per unit volume (or the ratio of pore volume to pore radius), to the fraction of waste containing solution, and to the inverse of the fraction of calcium in the saltstone. Rates and diffusivities are not proportional to the water to cement ratio, because this number depends on whether the fly ash is counted as cementitious, as in Class C cement, or not cementitious, as in Class F cement. In fact the relatively small amount of calcium in Class F cement contributes to the cementitious properties overall, though not so much as Class C cement. 4 refs., 2 figs., 6 tabs

Saltstone Osmotic Pressure

International Nuclear Information System (INIS)

Nichols, Ralph L.; Dixon, Kenneth L.

2013-01-01

groundwater will flow into the semi-permeable material resulting in hydrologic containment within the membrane. Additionally, hyperfiltration can occur within semi-permeable materials when water moves through a membrane into the more concentrated solution and dissolved constituents are retained in the lower concentration solution. Groundwater flow and transport equations that incorporate chemical gradients (osmosis) have been developed. These equations are referred to as coupled flow equations. Currently groundwater modeling to assess the performance of saltstone waste forms is conducted using the PORFLOW groundwater flow and transport model. PORFLOW does not include coupled flow from chemico-osmotic gradients and therefore numerical simulation of the effect of coupled flow on contaminant transport in and around saltstone cannot be assessed. Most natural semi-permeable membranes are non-ideal membranes and do not restrict all movement of solutes and as a result theoretical osmotic potential is not realized. Osmotic efficiency is a parameter in the coupled flow equation that accounts for the non-ideal behavior of most semi-permeable membranes. On order to evaluate the effects of osmotic potential on the hydraulic of a system the osmotic potential must be known. Several lab methods have been developed to measure osmotic efficiency for use in coupled flow analysis

Saltstone processing startup at the Savannah River Plant

International Nuclear Information System (INIS)

Wilhite, E.L.; Langton, C.A.; Sturm, H.F.; Hooker, R.L.; Occhipinti, E.S.

1988-01-01

High-level nuclear wastes are stored in large underground tanks at the Savannah River Plant. Processing of this waste in preparation for ultimate disposal will begin in 1988. The waste will be processed to separate the high-level radioactive fraction from the low-level radioactive fraction. The separation will be made in existing waste tanks by a process combining precipitation, adsorption, and filtration. The high-level fraction will be vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) for permanent disposal in a federal repository. The low-level fraction (decontaminated salt solution) will be mixed with a cementitious slag-flyash blend. The resulting wasteform, saltstone, will be disposed of onsite by emplacement in an engineered facility. Waste properties, disposal facility details, and wasteform characteristics are discussed. In particular, details of saltstone processing, focusing on experience obtained from facility startup, are presented

Degradation Of Cementitious Materials Associated With Saltstone Disposal Units

International Nuclear Information System (INIS)

Flach, G. P; Smith, F. G. III

2013-01-01

The Saltstone facilities at the DOE Savannah River Site (SRS) stabilize and dispose of low-level radioactive salt solution originating from liquid waste storage tanks at the site. The Saltstone Production Facility (SPF) receives treated salt solution and mixes the aqueous waste with dry cement, blast furnace slag, and fly ash to form a grout slurry which is mechanically pumped into concrete disposal cells that compose the Saltstone Disposal Facility (SDF). The solidified grout is termed ''saltstone''. Cementitious materials play a prominent role in the design and long-term performance of the SDF. The saltstone grout exhibits low permeability and diffusivity, and thus represents a physical barrier to waste release. The waste form is also reducing, which creates a chemical barrier to waste release for certain key radionuclides, notably Tc-99. Similarly, the concrete shell of an SDF disposal unit (SDU) represents an additional physical and chemical barrier to radionuclide release to the environment. Together the waste form and the SDU compose a robust containment structure at the time of facility closure. However, the physical and chemical state of cementitious materials will evolve over time through a variety of phenomena, leading to degraded barrier performance over Performance Assessment (PA) timescales of thousands to tens of thousands of years. Previous studies of cementitious material degradation in the context of low-level waste disposal have identified sulfate attack, carbonation influenced steel corrosion, and decalcification (primary constituent leaching) as the primary chemical degradation phenomena of most relevance to SRS exposure conditions. In this study, degradation time scales for each of these three degradation phenomena are estimated for saltstone and concrete associated with each SDU type under conservative, nominal, and best estimate assumptions. The nominal value (NV) is an intermediate result that is more probable than the conservative estimate

MEASUREMENT OF SPECIFIC HEAT CAPACITY OF SALTSTONE

International Nuclear Information System (INIS)

Harbour, J.; Williams, V.

2008-01-01

One of the goals of the Saltstone variability study is to identify (and quantify the impact of) the operational and compositional variables that control or influence the important processing and performance properties of Saltstone grout mixtures. The heat capacity of the Saltstone waste form is one of the important properties of Saltstone mixes that was last measured at SRNL in 1997. It is therefore important to develop a core competency for rapid and accurate analysis of the specific heat capacity of the Saltstone mixes in order to quantify the impact of compositional and operational variations on this property as part of the variability study. The heat capacity, coupled with the heat of hydration data obtained from isothermal calorimetry for a given Saltstone mix, can be used to predict the maximum temperature increase in the cells within the vaults of the Saltstone Disposal Facility (SDF). The temperature increase controls the processing rate and the pour schedule. The maximum temperature is also important to the performance properties of the Saltstone. For example, in mass pours of concrete or grout of which Saltstone is an example, the maximum temperature increase and the maximum temperature difference (between the surface and the hottest location) are controlled to ensure durability of the product and prevent or limit the cracking caused by the thermal gradients produced during curing. This report details the development and implementation of a method for the measurement of the heat capacities of Saltstone mixes as well as the heat capacities of the cementitious materials of the premix and the simulated salt solutions used to batch the mixes. The developed method utilizes the TAM Air isothermal calorimeter and takes advantage of the sophisticated heat flow measurement capabilities of the instrument. Standards and reference materials were identified and used to validate the procedure and ensure accuracy of testing. Heat capacities of Saltstone mixes were �

MEASUREMENT OF SPECIFIC HEAT CAPACITY OF SALTSTONE

Energy Technology Data Exchange (ETDEWEB)

Harbour, J; Vickie Williams, V

2008-09-29

One of the goals of the Saltstone variability study is to identify (and quantify the impact of) the operational and compositional variables that control or influence the important processing and performance properties of Saltstone grout mixtures. The heat capacity of the Saltstone waste form is one of the important properties of Saltstone mixes that was last measured at SRNL in 1997. It is therefore important to develop a core competency for rapid and accurate analysis of the specific heat capacity of the Saltstone mixes in order to quantify the impact of compositional and operational variations on this property as part of the variability study. The heat capacity, coupled with the heat of hydration data obtained from isothermal calorimetry for a given Saltstone mix, can be used to predict the maximum temperature increase in the cells within the vaults of the Saltstone Disposal Facility (SDF). The temperature increase controls the processing rate and the pour schedule. The maximum temperature is also important to the performance properties of the Saltstone. For example, in mass pours of concrete or grout of which Saltstone is an example, the maximum temperature increase and the maximum temperature difference (between the surface and the hottest location) are controlled to ensure durability of the product and prevent or limit the cracking caused by the thermal gradients produced during curing. This report details the development and implementation of a method for the measurement of the heat capacities of Saltstone mixes as well as the heat capacities of the cementitious materials of the premix and the simulated salt solutions used to batch the mixes. The developed method utilizes the TAM Air isothermal calorimeter and takes advantage of the sophisticated heat flow measurement capabilities of the instrument. Standards and reference materials were identified and used to validate the procedure and ensure accuracy of testing. Heat capacities of Saltstone mixes were

Characterization Of Core Sample Collected From The Saltstone Disposal Facility

International Nuclear Information System (INIS)

Cozzi, A.; Duncan, A.

2010-01-01

During the month of September 2008, grout core samples were collected from the Saltstone Disposal Facility, Vault 4, cell E. This grout was placed during processing campaigns in December 2007 from Deliquification, Dissolution and Adjustment Batch 2 salt solution. The 4QCY07 Waste Acceptance Criteria sample collected on 11/16/07 represents the salt solution in the core samples. Core samples were retrieved to initiate the historical database of properties of emplaced Saltstone and to demonstrate the correlation between field collected and laboratory prepared samples. Three samples were collected from three different locations. Samples were collected using a two-inch diameter concrete coring bit. In April 2009, the core samples were removed from the evacuated sample container, inspected, transferred to PVC containers, and backfilled with nitrogen. Samples furthest from the wall were the most intact cylindrically shaped cored samples. The shade of the core samples darkened as the depth of coring increased. Based on the visual inspection, sample 3-3 was selected for all subsequent analysis. The density and porosity of the Vault 4 core sample, 1.90 g/cm 3 and 59.90% respectively, were comparable to values achieved for laboratory prepared samples. X-ray diffraction analysis identified phases consistent with the expectations for hydrated Saltstone. Microscopic analysis revealed morphology features characteristic of cementitious materials with fly ash and calcium silicate hydrate gel. When taken together, the results of the density, porosity, x-ray diffraction analysis and microscopic analysis support the conclusion that the Vault 4, Cell E core sample is representative of the expected waste form.

Degradation Of Cementitious Materials Associated With Saltstone Disposal Units

Energy Technology Data Exchange (ETDEWEB)

Flach, G. P; Smith, F. G. III

2013-03-19

The Saltstone facilities at the DOE Savannah River Site (SRS) stabilize and dispose of low-level radioactive salt solution originating from liquid waste storage tanks at the site. The Saltstone Production Facility (SPF) receives treated salt solution and mixes the aqueous waste with dry cement, blast furnace slag, and fly ash to form a grout slurry which is mechanically pumped into concrete disposal cells that compose the Saltstone Disposal Facility (SDF). The solidified grout is termed “saltstone”. Cementitious materials play a prominent role in the design and long-term performance of the SDF. The saltstone grout exhibits low permeability and diffusivity, and thus represents a physical barrier to waste release. The waste form is also reducing, which creates a chemical barrier to waste release for certain key radionuclides, notably Tc-99. Similarly, the concrete shell of an SDF disposal unit (SDU) represents an additional physical and chemical barrier to radionuclide release to the environment. Together the waste form and the SDU compose a robust containment structure at the time of facility closure. However, the physical and chemical state of cementitious materials will evolve over time through a variety of phenomena, leading to degraded barrier performance over Performance Assessment (PA) timescales of thousands to tens of thousands of years. Previous studies of cementitious material degradation in the context of low-level waste disposal have identified sulfate attack, carbonation influenced steel corrosion, and decalcification (primary constituent leaching) as the primary chemical degradation phenomena of most relevance to SRS exposure conditions. In this study, degradation time scales for each of these three degradation phenomena are estimated for saltstone and concrete associated with each SDU type under conservative, nominal, and best estimate assumptions. The nominal value (NV) is an intermediate result that is more probable than the conservative

Leaching of saltstone: Laboratory and field testing and mathematical modeling

International Nuclear Information System (INIS)

Grant, M.W.; Langton, C.A.; Oblath, S.B.; Pepper, D.W.; Wallace, R.M.; Wilhite, E.L.; Yau, W.W.F.

1987-01-01

A low-level alkaline salt solution will be a byproduct in the processing of high-level waste at the Savannah River Plant (SRP). This solution will be incorporated into a wasteform, saltstone, and disposed of in surface vaults. Laboratory and field leach testing and mathematical modeling have demonstrated the predictability of contaminant release from cement wasteforms. Saltstone disposal in surface vaults will meet the design objective, which is to meet drinking water standards in shallow groundwater at the disposal area boundary. Diffusion is the predominant mechanism for release of contaminants to the environment. Leach testing in unsaturated soil, at soil moisture levels above 1 wt %, has shown no difference in leach rate compared to leaching in distilled water. Field leach testing of three thirty-ton blocks of saltstone in lysimeters has been underway since January 1984. Mathematical models were applied to assess design features for saltstone disposal. One dimensional infinite-composite and semi-infinite analytical models were developed for assessing diffusion of nitrate from saltstone through a cement barrier. Numerical models, both finite element and finite difference, were validated by comparison of model predictions with the saltstone lysimeter results. Validated models were used to assess the long-term performance of the saltstone stored in surface vaults. The maximum concentrations of all contaminants released from saltstone to shallow groundwater are predicted to be below drinking water standards at the disposal area boundary. 5 refs., 11 figs., 5 tabs

Benzene Evolution Rates from Saltstone Prepared with 2X ITP Flowsheet Concentrations of Phenylborates and Heated to 85 Degrees C

International Nuclear Information System (INIS)

Poirier, M.R.

2000-01-01

The Saltstone Facility provides the final treatment and disposal of low level liquid wastes streams. At the Saltstone Facility, the waste is mixed with cement, flyash, and slag to form a grout, which is pumped into large concrete vaults where it cures. The facility started radioactive operations in June 1990. High Level Waste Engineering requested Savannah River Technology Center to determine the effect of TPB and its decomposition products (i.e., 3PB, 2PB, and 1PB) on the saltstone process. Previous testing performed by SRTC determined saltstone benzene evolution rates a function of ITP filtrate composition. Testing by the Thermal Fluids Laboratory has shown at design operation, the temperature in the Z-area vaults could reach 85 degrees Celsius. Saltstone asked SRTC to perform additional testing to determine whether curing at 85 degrees Celsius could change saltstone benzene evolution rates. This document describes the test performed to determine the effect of curing temperature on the benzene evolution rates

Composite analysis E-area vaults and saltstone disposal facilities

Energy Technology Data Exchange (ETDEWEB)

Cook, J.R.

1997-09-01

This report documents the Composite Analysis (CA) performed on the two active Savannah River Site (SRS) low-level radioactive waste (LLW) disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults (EAV) Disposal Facility. The analysis calculated potential releases to the environment from all sources of residual radioactive material expected to remain in the General Separations Area (GSA). The GSA is the central part of SRS and contains all of the waste disposal facilities, chemical separations facilities and associated high-level waste storage facilities as well as numerous other sources of radioactive material. The analysis considered 114 potential sources of radioactive material containing 115 radionuclides. The results of the CA clearly indicate that continued disposal of low-level waste in the saltstone and EAV facilities, consistent with their respective radiological performance assessments, will have no adverse impact on future members of the public.

Composite analysis E-area vaults and saltstone disposal facilities

International Nuclear Information System (INIS)

Cook, J.R.

1997-09-01

This report documents the Composite Analysis (CA) performed on the two active Savannah River Site (SRS) low-level radioactive waste (LLW) disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults (EAV) Disposal Facility. The analysis calculated potential releases to the environment from all sources of residual radioactive material expected to remain in the General Separations Area (GSA). The GSA is the central part of SRS and contains all of the waste disposal facilities, chemical separations facilities and associated high-level waste storage facilities as well as numerous other sources of radioactive material. The analysis considered 114 potential sources of radioactive material containing 115 radionuclides. The results of the CA clearly indicate that continued disposal of low-level waste in the saltstone and EAV facilities, consistent with their respective radiological performance assessments, will have no adverse impact on future members of the public

SALTSTONE AND RADIONUCLIDE INTERACTIONS: RADIONUCLIDE SORPTION AND DESORPTION, AND SALTSTONE REDUCTION CAPACITY

International Nuclear Information System (INIS)

Kaplan, D; Roberts, Kimberly; Serkiz, Steven; Siegfried, Matthew

2008-01-01

The overall objective of this study was to measure a number of key input parameters quantifying geochemical processes in the subsurface environment of the Savannah River Site's (SRS's) Saltstone Facility. For the first time, sorption (K d ) values of numerous radionuclides were measured with Saltstone and Vault 2 concrete. Particular attention was directed at understanding how Tc adsorbs and desorbs from these cementitious materials with the intent to demonstrate that desorption occurs at a much slower rate than adsorption, thus permitting the use of kinetic terms instead of (or along with) the steady state K d term. Another very important parameter measured was the reduction capacity of these materials. This parameter is used to estimate the duration that the Saltstone facility remains in a reduced chemical state, a condition that maintains several otherwise mobile radionuclides in an immobile form. Key findings of this study follow. K d values for Am, Cd, Ce, Co, Cs, Hg, I, Np, Pa, Pu, Se, Sn, Tc, U, and Y for Saltstone and Vault 2 concrete were measured under oxidized and reduced conditions. Precipitation of several of the higher valence state radionuclides was observed. There was little evidence that the Vault 2 and Saltstone K d values differed from previous SRS K d values measured with reducing grout (Kaplan and Coates 2007). These values also supported a previous finding that K d values of slag-containing cementitious materials, tend to be greater for cations and about the same for anions, than regular cementitious materials without slag. Based on these new findings, it was suggested that all previous reducing concrete K d values be used in future PAs, except Np(V) and Pu(IV) K d values, which should be increased, and I values, which should be slightly decreased in all three stages of concrete aging. The reduction capacity of Saltstone, consisting of 23 wt-% blast furnace slag, was 821.8 microequivalents per gram ((micro)eq/g). This value was approximately

SALTSTONE AND RADIONUCLIDE INTERACTIONS: RADIONUCLIDE SORPTION AND DESORPTION, AND SALTSTONE REDUCTION CAPACITY

Energy Technology Data Exchange (ETDEWEB)

Kaplan, D; Kimberly Roberts, K; Steven Serkiz, S; Matthew Siegfried, M

2008-10-30

The overall objective of this study was to measure a number of key input parameters quantifying geochemical processes in the subsurface environment of the Savannah River Site's (SRS's) Saltstone Facility. For the first time, sorption (K{sub d}) values of numerous radionuclides were measured with Saltstone and Vault 2 concrete. Particular attention was directed at understanding how Tc adsorbs and desorbs from these cementitious materials with the intent to demonstrate that desorption occurs at a much slower rate than adsorption, thus permitting the use of kinetic terms instead of (or along with) the steady state K{sub d} term. Another very important parameter measured was the reduction capacity of these materials. This parameter is used to estimate the duration that the Saltstone facility remains in a reduced chemical state, a condition that maintains several otherwise mobile radionuclides in an immobile form. Key findings of this study follow. K{sub d} values for Am, Cd, Ce, Co, Cs, Hg, I, Np, Pa, Pu, Se, Sn, Tc, U, and Y for Saltstone and Vault 2 concrete were measured under oxidized and reduced conditions. Precipitation of several of the higher valence state radionuclides was observed. There was little evidence that the Vault 2 and Saltstone K{sub d} values differed from previous SRS K{sub d} values measured with reducing grout (Kaplan and Coates 2007). These values also supported a previous finding that K{sub d} values of slag-containing cementitious materials, tend to be greater for cations and about the same for anions, than regular cementitious materials without slag. Based on these new findings, it was suggested that all previous reducing concrete K{sub d} values be used in future PAs, except Np(V) and Pu(IV) K{sub d} values, which should be increased, and I values, which should be slightly decreased in all three stages of concrete aging. The reduction capacity of Saltstone, consisting of 23 wt-% blast furnace slag, was 821.8 microequivalents per

Technical Insights for Saltstone PA Maintenance

International Nuclear Information System (INIS)

Flach, G.; Sarkar, S.; Mahadevan, S.; Kosson, D.

2011-01-01

The Cementitious Barriers Partnership (CBP) is a collaborative program sponsored by the US DOE Office of Waste Processing. The objective of the CBP is to develop a set of computational tools to improve understanding and prediction of the long-term structural, hydraulic, and chemical performance of cementitious barriers and waste forms used in nuclear applications. CBP tools are expected to better characterize and reduce the uncertainties of current methodologies for assessing cementitious barrier performance and increase the consistency and transparency of the assessment process, as the five-year program progresses. In September 2009, entering its second year of funded effort, the CBP sought opportunities to provide near-term tangible support to DOE Performance Assessments (PAs). The Savannah River Saltstone Disposal Facility (SDF) was selected for the initial PA support effort because (1) cementitious waste forms and barriers play a prominent role in the performance of the facility, (2) certain important long-term behaviors of cementitious materials composing the facility are uncertain, (3) review of the SDF PA by external stakeholders is ongoing, and (4) the DOE contractor responsible for the SDF PA is open to receiving technical assistance from the CBP. A review of the current (SRR Closure and Waste Disposal Authority 2009) and prior Saltstone PAs (e.g., Cook et al. 2005) suggested five potential opportunities for improving predictions. The candidate topics considered were (1) concrete degradation from external sulfate attack, (2) impact of atmospheric exposure to concrete and grout before closure, such as accelerated slag and Tc-99 oxidation, (3) mechanistic prediction of geochemical conditions, (4) concrete degradation from rebar corrosion due to carbonation, and (5) early age cracking from drying and/or thermal shrinkage. The candidate topics were down-selected considering the feasibility of addressing each issue within approximately six months, and

Technical Insights for Saltstone PA Maintenance

Energy Technology Data Exchange (ETDEWEB)

Flach, G.; Sarkar, S.; Mahadevan, S.; Kosson, D.

2011-07-20

The Cementitious Barriers Partnership (CBP) is a collaborative program sponsored by the US DOE Office of Waste Processing. The objective of the CBP is to develop a set of computational tools to improve understanding and prediction of the long-term structural, hydraulic, and chemical performance of cementitious barriers and waste forms used in nuclear applications. CBP tools are expected to better characterize and reduce the uncertainties of current methodologies for assessing cementitious barrier performance and increase the consistency and transparency of the assessment process, as the five-year program progresses. In September 2009, entering its second year of funded effort, the CBP sought opportunities to provide near-term tangible support to DOE Performance Assessments (PAs). The Savannah River Saltstone Disposal Facility (SDF) was selected for the initial PA support effort because (1) cementitious waste forms and barriers play a prominent role in the performance of the facility, (2) certain important long-term behaviors of cementitious materials composing the facility are uncertain, (3) review of the SDF PA by external stakeholders is ongoing, and (4) the DOE contractor responsible for the SDF PA is open to receiving technical assistance from the CBP. A review of the current (SRR Closure & Waste Disposal Authority 2009) and prior Saltstone PAs (e.g., Cook et al. 2005) suggested five potential opportunities for improving predictions. The candidate topics considered were (1) concrete degradation from external sulfate attack, (2) impact of atmospheric exposure to concrete and grout before closure, such as accelerated slag and Tc-99 oxidation, (3) mechanistic prediction of geochemical conditions, (4) concrete degradation from rebar corrosion due to carbonation, and (5) early age cracking from drying and/or thermal shrinkage. The candidate topics were down-selected considering the feasibility of addressing each issue within approximately six months, and

Groundwater Monitoring Plan for the Z-Area Saltstone Facility

International Nuclear Information System (INIS)

Wells, D.

2002-01-01

Groundwater monitoring has been conducted at the Z-Area Saltstone Disposal Facility since 1987. At that time, groundwater monitoring was not required by the industrial landfill regulations, but a modest monitoring program was required by the operating permit. In 1996 SRS proposed a program based on direct push sampling. This program called for biennial direct push sampling within 25 feet of each waste-containing cell with additional samples being taken in areas where excessive cracking had been observed. The direct push proposal was accepted by The South Carolina Department of Health and Environmental Control (SCDHEC). The Industrial Solid Waste Landfill Regulations were revised in 1998 and now include requirements for groundwater monitoring. The major elements of those regulations and their application at Z-Area are discussed. These are a point of compliance, groundwater protection standards, the groundwater monitoring system, sampling and analysis, and data evaluation and reporting

Saltstone Osmotic Pressure

Energy Technology Data Exchange (ETDEWEB)

Nichols, Ralph L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, Kenneth L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRN

2013-09-23

Recent research into the moisture retention properties of saltstone suggest that osmotic pressure may play a potentially significant role in contaminant transport (Dixon et al., 2009 and Dixon, 2011). The Savannah River Remediation Closure and Disposal Assessments Group requested the Savannah River National Laboratory (SRNL) to conduct a literature search on osmotic potential as it relates to contaminant transport and to develop a conceptual model of saltstone that incorporates osmotic potential. This report presents the findings of the literature review and presents a conceptual model for saltstone that incorporates osmotic potential. The task was requested through Task Technical Request HLW-SSF-TTR- 2013-0004.

Distribution Coeficients (Kd) Generated From A Core Sample Collected From The Saltstone Disposal Facility

International Nuclear Information System (INIS)

Almond, P.; Kaplan, D.

2011-01-01

Core samples originating from Vault 4, Cell E of the Saltstone Disposal Facility (SDF) were collected in September of 2008 (Hansen and Crawford 2009, Smith 2008) and sent to SRNL to measure chemical and physical properties of the material including visual uniformity, mineralogy, microstructure, density, porosity, distribution coefficients (K d ), and chemical composition. Some data from these experiments have been reported (Cozzi and Duncan 2010). In this study, leaching experiments were conducted with a single core sample under conditions that are representative of saltstone performance. In separate experiments, reducing and oxidizing environments were targeted to obtain solubility and Kd values from the measurable species identified in the solid and aqueous leachate. This study was designed to provide insight into how readily species immobilized in saltstone will leach from the saltstone under oxidizing conditions simulating the edge of a saltstone monolith and under reducing conditions, targeting conditions within the saltstone monolith. Core samples were taken from saltstone poured in December of 2007 giving a cure time of nine months in the cell and a total of thirty months before leaching experiments began in June 2010. The saltstone from Vault 4, Cell E is comprised of blast furnace slag, class F fly ash, portland cement, and Deliquification, Dissolution, and Adjustment (DDA) Batch 2 salt solution. The salt solution was previously analyzed from a sample of Tank 50 salt solution and characterized in the 4QCY07 Waste Acceptance Criteria (WAC) report (Zeigler and Bibler 2009). Subsequent to Tank 50 analysis, additional solution was added to the tank solution from the Effluent Treatment Project as well as from inleakage from Tank 50 pump bearings (Cozzi and Duncan 2010). Core samples were taken from three locations and at three depths at each location using a two-inch diameter concrete coring bit (1-1, 1-2, 1-3; 2-1, 2-2, 2-3; 3-1, 3-2, 3-3) (Hansen and Crawford

DISTRIBUTION COEFICIENTS (KD) GENERATED FROM A CORE SAMPLE COLLECTED FROM THE SALTSTONE DISPOSAL FACILITY

Energy Technology Data Exchange (ETDEWEB)

Almond, P.; Kaplan, D.

2011-04-25

Core samples originating from Vault 4, Cell E of the Saltstone Disposal Facility (SDF) were collected in September of 2008 (Hansen and Crawford 2009, Smith 2008) and sent to SRNL to measure chemical and physical properties of the material including visual uniformity, mineralogy, microstructure, density, porosity, distribution coefficients (K{sub d}), and chemical composition. Some data from these experiments have been reported (Cozzi and Duncan 2010). In this study, leaching experiments were conducted with a single core sample under conditions that are representative of saltstone performance. In separate experiments, reducing and oxidizing environments were targeted to obtain solubility and Kd values from the measurable species identified in the solid and aqueous leachate. This study was designed to provide insight into how readily species immobilized in saltstone will leach from the saltstone under oxidizing conditions simulating the edge of a saltstone monolith and under reducing conditions, targeting conditions within the saltstone monolith. Core samples were taken from saltstone poured in December of 2007 giving a cure time of nine months in the cell and a total of thirty months before leaching experiments began in June 2010. The saltstone from Vault 4, Cell E is comprised of blast furnace slag, class F fly ash, portland cement, and Deliquification, Dissolution, and Adjustment (DDA) Batch 2 salt solution. The salt solution was previously analyzed from a sample of Tank 50 salt solution and characterized in the 4QCY07 Waste Acceptance Criteria (WAC) report (Zeigler and Bibler 2009). Subsequent to Tank 50 analysis, additional solution was added to the tank solution from the Effluent Treatment Project as well as from inleakage from Tank 50 pump bearings (Cozzi and Duncan 2010). Core samples were taken from three locations and at three depths at each location using a two-inch diameter concrete coring bit (1-1, 1-2, 1-3; 2-1, 2-2, 2-3; 3-1, 3-2, 3-3) (Hansen and

Lysimeter study of vegetative uptake from saltstone. Part I. Design, installation, and data collection plan

International Nuclear Information System (INIS)

Johnson, T.L.

1986-02-01

A field test facility has been designed and installed to obtain data on the vegetative uptake of radionuclides from buried low-level radioactive waste. The waste is a cement-like, solidified salt solution known as saltstone. The facility consists of 32 lysimeters (containers 6 feet in diameter and 6 to 10 feet in depth) holding buried saltstone at varying depths, and with varying types of vegetation grown at the surface. Vegetation, soil, and groundwater samples will be analyzed for Tc-99, Sr-90, I-129, Cs-137, and other radionuclides. Groundwater will also be analyzed for other water quality parameters, including nitrates

Design of saltstone vaults

International Nuclear Information System (INIS)

Aiyar, G.S.; Hsiu, F.J.

1987-01-01

Radioactive waste from processed spent nuclear fuel at the Savannah River Plant, South Carolina, are stored in underground tanks. The wastes consist of sludge and supernate. Most of the radionuclides and some nonradioactive constituents are removed from the supernate. These along with the sludge are converted into a glass-crystallite matrix and cast into stainless steel cansisters for future disposal in a geological repository. The decontaminated salt solution is mixed with cement, fly ash, and a set-retarding agent, and the resulting grout is transferred to reinforced concrete vaults where it sets into a monolith termed saltstone. The vault is then capped with concrete. A total of 21 vaults measuring 600 x 100 x 27 ft are planned for disposal of the existing supernate plus any additional supernate generated up to the year 2000

Results and analysis of saltstone cores taken from saltstone disposal unit cell 2A

Energy Technology Data Exchange (ETDEWEB)

Reigel, M. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-03-01

As part of an ongoing Performance Assessment (PA) Maintenance Plan, Savannah River Remediation (SRR) has developed a sampling and analyses strategy to facilitate the comparison of field-emplaced samples (i.e., saltstone placed and cured in a Saltstone Disposal Unit (SDU)) with samples prepared and cured in the laboratory. The primary objectives of the Sampling and Analyses Plan (SAP) are; (1) to demonstrate a correlation between the measured properties of laboratory-prepared, simulant samples (termed Sample Set 3), and the field-emplaced saltstone samples (termed Sample Set 9), and (2) to validate property values assumed for the Saltstone Disposal Facility (SDF) PA modeling. The analysis and property data for Sample Set 9 (i.e. six core samples extracted from SDU Cell 2A (SDU2A)) are documented in this report, and where applicable, the results are compared to the results for Sample Set 3. Relevant properties to demonstrate the aforementioned objectives include bulk density, porosity, saturated hydraulic conductivity (SHC), and radionuclide leaching behavior.

Development and Implementation of a Scaled Saltstone Facility at Savannah River National Laboratory - 13346

International Nuclear Information System (INIS)

Reigel, Marissa M.; Fowley, Mark D.; Hansen, Erich K.; Hera, Kevin R.; Marzolf, Athneal D.; Cozzi, Alex D.

2013-01-01

The Savannah River National Laboratory (SRNL) has supported the Saltstone Production Facility (SPF) since its conception. However, bench scaled tests have not always provided process or performance data related to the mixing, transfer, and other operations utilized in the SPF. A need was identified to better understand the SPF processes and to have the capabilities at SRNL to simulate the SPF unit operations to support an active low-level radioactive waste (LLW) processing facility. At the SPF, the dry premix is weighed, mixed and transferred to the Readco '10-inch' continuous mixer where it is mixed with the LLW salt solution from the Salt Feed Tank (SFT) to produce fresh Saltstone slurry. The slurry is discharged from the mixer into a hopper. The hopper feeds the grout pump that transfers the slurry through at least 457.2 meters of piping and discharges it into the Saltstone Disposal Units (SDU) for permanent disposal. In conjunction with testing individual SPF processes over several years, SRNL has designed and fabricated a scaled Saltstone Facility. Scaling of the system is primarily based on the volume capacity of the mixer and maintaining the same shear rate and total shear at the wall of the transfer line. At present, SRNL is utilizing the modular capabilities of the scaled Saltstone Facility to investigate the erosion issues related to the augers and paddles inside the SPF mixer. Full implementation of the scaled Saltstone Facility is still ongoing, but it is proving to be a valuable resource for testing alternate Saltstone formulations, cleaning sequences, the effect of pumping Saltstone to farther SDU's, optimization of the SPF mixer, and other operational variables before they are implemented in the SPF. (authors)

Performance Properties Of Saltstone Produced Using SWPF Simulants

International Nuclear Information System (INIS)

Harbour, J.; Edwards, T.

2010-01-01

The overwhelming majority of waste to be immobilized at the Saltstone Production Facility will come from the waste stream exiting the Salt Waste Processing Facility (SWPF). These SWPF batches are salt solutions that result from pretreatment of the High Level Waste (HLW) supernate by an Actinide Removal Process followed by Caustic Side Solvent Extraction. The concentration of aluminate within these streams will vary and be determined by (1) the concentration in the incoming salt waste stream, (2) the degree of aluminum leaching from the HLW, (3) the method for introducing the aluminate into the waste stream (continuous or batch) and (4) and any operational or regulatory limitations. The overall Performance Assessment outcome for the Saltstone Disposal Facility will depend significantly on the performance properties of the SWPF Saltstone grouts. This report identifies and quantifies, when possible, those factors that drive the performance properties of the projected SWPF grouts. Previous work has identified aluminate concentration in the salt waste stream as a key factor in determining performance. Consequently, significant variation in the aluminate concentration to a maximum level of 0.65 M was investigated in this report. The SWPF baseline grout is a mix with a 0.60 water to cementitious ratio and a premix composition of 45 wt % slag, 45 wt % fly ash and 10 wt % portland cement. The key factors that drive performance of the SWPF mixes were determined to be (1) the time/temperature profile for curing, (2) water to cementitious materials ratio, (3) aluminate concentration in the waste stream, and (4) wt % slag in the premix. An increase in the curing temperature for mixes with 45 wt % slag resulted in a 2.5 times decrease in Young's modulus. The reduction of Young's modulus measured at 60 C versus 22 C was mitigated by an increase in the aluminate concentration but was still significant. For mixes containing 60 wt % slag, the reduction in Young's modulus between

Literature Review of the Effects of Tetraphenylborate on Saltstone Grout: Benzene Evolution and TCLP Performance

International Nuclear Information System (INIS)

HAY, MICHAEL

2004-01-01

As part of the program to disposition the tetraphenylborate (TPB) in Tank 48H and return the tank to service, Salt Processing Development requested a review of the literature to assess the state of knowledge pertaining to incorporation of tetraphenylborate slurries in saltstone grout with respect to benzene generation rates and leaching performance. Examination of past studies conducted at Savannah River Site (SRS) on the incorporation of TPB slurries in saltstone provides a basis for developing a more focused scope of experimental studies. Tank 48H currently contains potassium and cesium tetraphenylborate salts as a result of a demonstration of the In Tank Precipitation (ITP) process in 1983 and subsequent ITPradioactive start-up operations in 1995. The tank currently contains approximately 240,000 gallons of salt solution with approximately 19,000 kg of potassium and cesium tetraphenylborate salts. The presence of the TPB salts makes the waste incompatible with existing High Level Waste treatment facilities. The TPB salts in Tank 48H must be treated or removed to meet the scheduled return to service date of 2007. The two preferred options for disposition of the TBP slurries in Tank 48H include: (1) Aggregation of the material with the Defense Waste Processing Facility (DWPF) recycle stream and disposal in the Saltstone Processing Facility (SPF), and (2) In-Situ Thermal Decomposition using heat in combination with pH reduction and catalyst addition. The current literature review along with the current experimental studies provide a basis for determining the feasibility of the option to incorporate the TPB slurries into saltstone grout

Saltstone Matrix Characterization And Stadium Simulation Results

International Nuclear Information System (INIS)

Langton, C.

2009-01-01

SIMCO Technologies, Inc. was contracted to evaluate the durability of the saltstone matrix material and to measure saltstone transport properties. This information will be used to: (1) Parameterize the STADIUM(reg s ign) service life code, (2) Predict the leach rate (degradation rate) for the saltstone matrix over 10,000 years using the STADIUM(reg s ign) concrete service life code, and (3) Validate the modeled results by conducting leaching (water immersion) tests. Saltstone durability for this evaluation is limited to changes in the matrix itself and does not include changes in the chemical speciation of the contaminants in the saltstone. This report summarized results obtained to date which include: characterization data for saltstone cured up to 365 days and characterization of saltstone cured for 137 days and immersed in water for 31 days. Chemicals for preparing simulated non-radioactive salt solution were obtained from chemical suppliers. The saltstone slurry was mixed according to directions provided by SRNL. However SIMCO Technologies Inc. personnel made a mistake in the premix proportions. The formulation SIMCO personnel used to prepare saltstone premix was not the reference mix proportions: 45 wt% slag, 45 wt% fly ash, and 10 wt% cement. SIMCO Technologies Inc. personnel used the following proportions: 21 wt% slag, 65 wt% fly ash, and 14 wt% cement. The mistake was acknowledged and new mixes have been prepared and are curing. The results presented in this report are assumed to be conservative since the excessive fly ash was used in the SIMCO saltstone. The SIMCO mixes are low in slag which is very reactive in the caustic salt solution. The impact is that the results presented in this report are expected to be conservative since the samples prepared were deficient in slag and contained excess fly ash. The hydraulic reactivity of slag is about four times that of fly ash so the amount of hydrated binder formed per unit volume in the SIMCO saltstone samples

Data Package for Secondary Waste Form Down-Selection-Cast Stone

International Nuclear Information System (INIS)

Serne, R. Jeffrey; Westsik, Joseph H.

2011-01-01

Available literature on Cast Stone and Saltstone was reviewed with an emphasis on determining how Cast Stone and related grout waste forms performed in relationship to various criteria that will be used to decide whether a specific type of waste form meets acceptance criteria for disposal in the Integrated Disposal Facility (IDF) at Hanford. After the critical review of the Cast Stone/Saltstone literature, we conclude that Cast Stone is a good candidate waste form for further consideration. Cast stone meets the target IDF acceptance criteria for compressive strength, no free liquids, TCLP leachate are below the UTS permissible concentrations and leach rates for Na and Tc-99 are suiteably low. The cost of starting ingredients and equipment necessary to generate Cast Stone waste forms with secondary waste streams are low and the Cast Stone dry blend formulation can be tailored to accommodate variations in liquid waste stream compositions. The database for Cast Stone short-term performance is quite extensive compared to the other three candidate waste solidification processes. The solidification of liquid wastes in Cast Stone is a mature process in comparison to the other three candidates. Successful production of Cast Stone or Saltstone has been demonstrated from lab-scale monoliths with volumes of cm3 through m3 sized blocks to 210-liter sized drums all the way to the large pours into vaults at Savannah River. To date over 9 million gallons of low activity liquid waste has been solidified and disposed in concrete vaults at Savannah River.

Data Package for Secondary Waste Form Down-Selection—Cast Stone

Energy Technology Data Exchange (ETDEWEB)

Serne, R. Jeffrey; Westsik, Joseph H.

2011-09-05

Available literature on Cast Stone and Saltstone was reviewed with an emphasis on determining how Cast Stone and related grout waste forms performed in relationship to various criteria that will be used to decide whether a specific type of waste form meets acceptance criteria for disposal in the Integrated Disposal Facility (IDF) at Hanford. After the critical review of the Cast Stone/Saltstone literature, we conclude that Cast Stone is a good candidate waste form for further consideration. Cast stone meets the target IDF acceptance criteria for compressive strength, no free liquids, TCLP leachate are below the UTS permissible concentrations and leach rates for Na and Tc-99 are suiteably low. The cost of starting ingredients and equipment necessary to generate Cast Stone waste forms with secondary waste streams are low and the Cast Stone dry blend formulation can be tailored to accommodate variations in liquid waste stream compositions. The database for Cast Stone short-term performance is quite extensive compared to the other three candidate waste solidification processes. The solidification of liquid wastes in Cast Stone is a mature process in comparison to the other three candidates. Successful production of Cast Stone or Saltstone has been demonstrated from lab-scale monoliths with volumes of cm3 through m3 sized blocks to 210-liter sized drums all the way to the large pours into vaults at Savannah River. To date over 9 million gallons of low activity liquid waste has been solidified and disposed in concrete vaults at Savannah River.

Special Analysis: Revision of Saltstone Vault 4 Disposal Limits (U)

Energy Technology Data Exchange (ETDEWEB)

Cook, J

2005-05-26

New disposal limits have been computed for Vault 4 of the Saltstone Disposal Facility based on several revisions to the models in the existing Performance Assessment and the Special Analysis issued in 2002. The most important changes are the use of a more rigorous groundwater flow and transport model, and consideration of radon emanation. Other revisions include refinement of the aquifer mesh to more accurately model the footprint of the vault, a new plutonium chemistry model accounting for the different transport properties of oxidation states III/IV and V/VI, use of variable infiltration rates to simulate degradation of the closure system, explicit calculation of gaseous releases and consideration of the effects of settlement and seismic activity on the vault structure. The disposal limits have been compared with the projected total inventory expected to be disposed in Vault 4. The resulting sum-of-fractions of the 1000-year disposal limits is 0.2, which indicates that the performance objectives and requirements of DOE 435.1 will not be exceeded. This SA has not altered the conceptual model (i.e., migration of radionuclides from the Saltstone waste form and Vault 4 to the environment via the processes of diffusion and advection) of the Saltstone PA (MMES 1992) nor has it altered the conclusions of the PA (i.e., disposal of the proposed waste in the SDF will meet DOE performance measures). Thus a PA revision is not required and this SA serves to update the disposal limits for Vault 4. In addition, projected doses have been calculated for comparison with the performance objectives laid out in 10 CFR 61. These doses are 0.05 mrem/year to a member of the public and 21.5 mrem/year to an inadvertent intruder in the resident scenario over a 10,000-year time-frame, which demonstrates that the 10 CFR 61 performance objectives will not be exceeded. This SA supplements the Saltstone PA and supersedes the two previous SAs (Cook et al. 2002; Cook and Kaplan 2003).

Addendum to the composite analysis for the E-Area Vaults and Saltstone Disposal Facilities

Energy Technology Data Exchange (ETDEWEB)

Cook, J.R.

2000-03-13

This report documents the composite analysis performed on the two active SRS low-level radioactive waste disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults Disposal Facility.

Addendum to the composite analysis for the E-Area Vaults and Saltstone Disposal Facilities

International Nuclear Information System (INIS)

Cook, J.R.

2000-01-01

This report documents the composite analysis performed on the two active SRS low-level radioactive waste disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults Disposal Facility

Delisting petition for 300-M saltstone (treated F006 sludge) from the 300-M liquid effluent treatment facility

Energy Technology Data Exchange (ETDEWEB)

1989-04-04

This petition seeks exclusion for stabilized and solidified sludge material generated by treatment of wastewater from the 300-M aluminum forming and metal finishing processes. The waste contains both hazardous and radioactive components and is classified as a mixed waste. The objective of this petition is to demonstrate that the stabilized sludge material (saltstone), when properly disposed, will not exceed the health-based standards for the hazardous constituents. This petition contains sampling and analytical data which justify the request for exclusion. The results show that when the data are applied to the EPA Vertical and Horizontal Spread (VHS) Model, health-based standards for all hazardous waste constituents will not be exceeded during worst case operating and environmental conditions. Disposal of the stabilized sludge material in concrete vaults will meet the requirements pertaining to Waste Management Activities for Groundwater Protection at the Savannah River Site in Aiken, S.C. Documents set forth performance objectives and disposal options for low-level radioactive waste disposal. Concrete vaults specified for disposal of 300-M saltstone (treated F006 sludge) assure that these performance objectives will be met.

Process Formulations And Curing Conditions That Affect Saltstone Properties

Energy Technology Data Exchange (ETDEWEB)

Reigel, M. M.; Pickenheim, B. R.; Daniel, W. E.

2012-09-28

The first objective of this study was to analyze saltstone fresh properties to determine the feasibility of reducing the formulation water to premix (w/p) ratio while varying the amount of extra water and admixtures used during processing at the Saltstone Production Facility (SPF). The second part of this study was to provide information for understanding the impact of curing conditions (cure temperature, relative humidity (RH)) and processing formulation on the performance properties of cured saltstone.

Key Factors That Influence The Performance Properties Of ARP/MCU Saltstone Mixes

International Nuclear Information System (INIS)

Harbour, J.; Edwards, T.; Williams, V.

2009-01-01

At the Saltstone Production Facility (SPF), decontaminated salt solution (DSS) is combined with premix (a cementitious mixture of portland cement (PC), blast furnace slag (BFS) and Class F fly ash (FA)) in a Readco mixer to produce fresh (uncured) Saltstone. After transfer to the Saltstone Disposal Facility (SDF) the hydration reactions initiated during the contact of the premix and salt solution continue during the curing period to produce the hardened waste form product. The amount of heat generated from hydration and the resultant temperature increase in the vaults depend on the composition of the decontaminated salt solution being dispositioned as well as the grout formulation (mix design). This report details the results from Task 3 of the Saltstone Variability Study for FY09 which was performed to identify, and quantify when possible, those factors that drive the performance properties of the projected ARP/MCU Batches. A baseline ARP/MCU mix (at 0.60 water to cementitious materials (w/cm) ratio) was established and consisted of the normal premix composition and a salt solution that was an average of the projected compositions of the last three ARP/MCU batches developed by T. A. Le. This task introduced significant variation in (1) wt % slag, w/cm ratio, and wt % portland cement about the baseline mix and (2) the temperature of curing in order to better assess the dependence of the performance properties on these factors. Two separate campaigns, designated Phase 10 and Phase 11, were carried out under Task 3. Experimental designs and statistical analyses were used to search for correlation among properties and to develop linear models to predict property values based on factors such as w/cm ratio, slag concentration, and portland cement concentration. It turns out that the projected salt compositions contained relatively high amounts of aluminate (0.22 M) even though no aluminate was introduced due to caustic aluminate removal from High Level Waste. Previous

Z-Area Saltstone Disposal Facility Groundwater Monitoring Report. 1997 Annual Report

International Nuclear Information System (INIS)

Roach, J.L. Jr.

1997-12-01

Samples from the ZBG wells at the Z-Area Saltstone Disposal Facility are analyzed for constituents required by South Carolina Department of Health and Environmental Control (SCDHEC) Industrial Solid Waste Permit number-sign 025500-1603 (formerly IWP-217). No constituents were reported above SCDHEC-proposed groundwater monitoring standards or final Primary Drinking Water Standards during first or third quareters 1997. No constituents were detected above SRS flagging criteria during first or third quarters 1997

SALTSTONE VARIABILITY STUDY - MEASUREMENT OF POROSITY

International Nuclear Information System (INIS)

Harbour, J; Vickie Williams, V; Tommy Edwards, T; Russell Eibling, R; Ray Schumacher, R

2007-01-01

One of the goals of the Saltstone Variability Study is to identify the operational and compositional variables that control or influence the important processing and performance properties of Saltstone mixes. One of the key performance properties is porosity which is a measure of the volume percent of a cured grout that is occupied by salt solution (for the saturated case). This report presents (1) the results of efforts to develop a method for the measurement of porosity of grout samples and (2) initial results of porosity values for samples that have been previously produced as part of the Saltstone Variability Study. A cost effective measurement method for porosity was developed that provides reproducible results, is relatively fast (30 to 60 minutes per sample) and uses a Mettler Toledo HR83 Moisture Analyzer that is already operational and routinely calibrated at Aiken County Technology Laboratory. The method involves the heating of the sample at 105 C until no further mass loss is observed. This mass loss value, which is due to water evaporation, is then used to calculate the volume percent porosity of the mix. The results of mass loss for mixes at 105 C were equivalent to the results obtained using thermal gravimetric analysis. The method was validated by comparing measurements of mass loss at 105 C for cured portland cement in water mixes to values presented in the literature for this system. A stereopycnometer from Quantachrome Instruments was selected to measure the cured grout bulk densities. Density is a property that is required to calculate the porosities. A stereopycnometer was already operational at Aiken County Technology Laboratory, has been calibrated using a solid stainless steel sphere of known volume, is cost effective and fast (∼15 minutes per sample). Cured grout densities are important in their own right because they can be used to project the volume of waste form produced from a given amount of salt feed of known composition. For mixes

Saltstone studies using the scaled continuous processing facility

Energy Technology Data Exchange (ETDEWEB)

Fowley, M. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hansen, E. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-08-01

The Savannah River National Laboratory (SRNL) has supported the Saltstone Facility since its conception with bench-scale laboratory experiments, mid-scale testing at vendor facilities, and consultations and testing at the Saltstone Facility. There have been minimal opportunities for the measurement of rheological properties of the grout slurry at the Saltstone Production Facility (SPF); thus, the Scaled Continuous Processing Facility (SCPF), constructed to provide processing data related to mixing, transfer, and other operations conducted in the SPF, is the most representative process data for determining the expected rheological properties in the SPF. These results can be used to verify the laboratory scale experiments that support the SPF using conventional mixing processes that appropriately represent the shear imparted to the slurry in the SPF.

GAS MIXING ANALYSIS IN A LARGE-SCALED SALTSTONE FACILITY

Energy Technology Data Exchange (ETDEWEB)

Lee, S

2008-05-28

Computational fluid dynamics (CFD) methods have been used to estimate the flow patterns mainly driven by temperature gradients inside vapor space in a large-scaled Saltstone vault facility at Savannah River site (SRS). The purpose of this work is to examine the gas motions inside the vapor space under the current vault configurations by taking a three-dimensional transient momentum-energy coupled approach for the vapor space domain of the vault. The modeling calculations were based on prototypic vault geometry and expected normal operating conditions as defined by Waste Solidification Engineering. The modeling analysis was focused on the air flow patterns near the ventilated corner zones of the vapor space inside the Saltstone vault. The turbulence behavior and natural convection mechanism used in the present model were benchmarked against the literature information and theoretical results. The verified model was applied to the Saltstone vault geometry for the transient assessment of the air flow patterns inside the vapor space of the vault region using the potential operating conditions. The baseline model considered two cases for the estimations of the flow patterns within the vapor space. One is the reference nominal case. The other is for the negative temperature gradient between the roof inner and top grout surface temperatures intended for the potential bounding condition. The flow patterns of the vapor space calculated by the CFD model demonstrate that the ambient air comes into the vapor space of the vault through the lower-end ventilation hole, and it gets heated up by the Benard-cell type circulation before leaving the vault via the higher-end ventilation hole. The calculated results are consistent with the literature information. Detailed results and the cases considered in the calculations will be discussed here.

IMPACT OF INCREASED ALUMINATE CONCENTRATIONS ON PROPERTIES OF SALTSTONE MIXES

International Nuclear Information System (INIS)

Harbour, J; Tommy Edwards, T; Erich Hansen, E; Vickie Williams, V

2007-01-01

One of the goals of the Saltstone variability study is to identify the operational and compositional variables that control or influence the important processing and performance properties of Saltstone mixes. The protocols developed in this variability study are ideally suited as a tool to assess the impact of proposed changes to the processing flow sheet for Liquid Waste Operations (LWO). One such proposal that is currently under consideration is to introduce a leaching step in the treatment of the High Level Waste (HLW) sludge to remove aluminum prior to vitrification at the Defense Waste Processing Facility (DWPF). This leachate would significantly increase the soluble aluminate concentrations as well as the free hydroxide ion concentration in the salt feed that will be processed at the Saltstone Processing Facility (SPF). Consequently, an initial study of the impact of increased aluminate concentration on the Saltstone grout properties was performed. The projected compositions and ranges of the aluminate rich salt stream (which includes the blending strategy) are not yet available and consequently, in this initial report, two separate salt stream compositions were investigated. The first stream starts with the previously projected baseline composition of the salt solution that will be fed to SPF from the Salt Waste Processing Facility (SWPF). The second stream is the solution that results from washing of the current Tank 51 sludge and subsequent transfer of the salt solution to Tank 11. The SWPF simulant has higher nitrate and lower free hydroxide than the Tank 11 simulant. In both of these cases, the aluminate was varied up to a maximum of 0.40 to 0.45M aluminate in order to evaluate the impact of increasing aluminate ion concentration on the grout properties. In general, the fresh grout properties of mixes made with SWPF and Tank 11 simulants were relatively insensitive to an increase in aluminate concentration in the salt solutions. However, the overall

Computational Fluid Dynamics Model for Saltstone Vault 4 Vapor Space

International Nuclear Information System (INIS)

Lee, Si Young

2005-01-01

Computational fluid dynamics (CFD) methods have been used to estimate the flow patterns for vapor space inside the Saltstone Vault No.4 under different operating scenarios. The purpose of this work is to examine the gas motions inside the vapor space under the current vault configurations. A CFD model took three-dimensional transient momentum-energy coupled approach for the vapor space domain of the vault. The modeling calculations were based on prototypic vault geometry and expected normal operating conditions as defined by Waste Solidification Engineering. The modeling analysis was focused on the air flow patterns near the ventilated corner zones of the vapor space inside the Saltstone vault. The turbulence behavior and natural convection mechanism used in the present model were benchmarked against the literature information and theoretical results. The verified model was applied to the Saltstone vault geometry for the transient assessment of the air flow patterns inside the vapor space of the vault region using the boundary conditions as provided by the customer. The present model considered two cases for the estimations of the flow patterns within the vapor space. One is the reference baseline case. The other is for the negative temperature gradient between the roof inner and top grout surface temperatures intended for the potential bounding condition. The flow patterns of the vapor space calculated by the CFD model demonstrate that the ambient air comes into the vapor space of the vault through the lower-end ventilation hole, and it gets heated up by the Benard-cell type circulation before leaving the vault via the higher-end ventilation hole. The calculated results are consistent with the literature information

Results for the second quarter 2014 tank 50 WAC slurry sample chemical and radionuclide contaminants

International Nuclear Information System (INIS)

Bannochie, C.

2014-01-01

This report details the chemical and radionuclide contaminant results for the characterization of the 2014 Second Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time. Information from this characterization will be used by DWPF & Saltstone Facility Engineering (DSFE) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System

Results for the Third Quarter 2014 Tank 50 WAC slurry sample: Chemical and radionuclide contaminants

Energy Technology Data Exchange (ETDEWEB)

Crawford, Charles L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-01-08

This report details the chemical and radionuclide contaminant results for the characterization of the 2014 Third Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time.¹ Information from this characterization will be used by DWPF & Saltstone Facility Engineering (DSFE) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System.

Results For The Fourth Quarter 2014 Tank 50 WAC Slurry Sample: Chemical And Radionuclide Contaminants

Energy Technology Data Exchange (ETDEWEB)

Crawford, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-09-30

This report details the chemical and radionuclide contaminant results for the characterization of the Calendar Year (CY) 2014 Fourth Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time. Information from this characterization will be used by DWPF & Saltstone Facility Engineering (DSFE) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System.

Results For The Third Quarter 2013 Tank 50 WAC Slurry Sample

Energy Technology Data Exchange (ETDEWEB)

Bannochie, Christopher J.

2013-11-26

This report details the chemical and radionuclide contaminant results for the characterization of the 2013 Third Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time. Information from this characterization will be used by DWPF & Saltstone Facility Engineering (DSFE) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System.

Results For The Second Quarter 2013 Tank 50 WAC Slurry Sample: Chemical And Radionuclide Contaminants

Energy Technology Data Exchange (ETDEWEB)

Bannochie, Christopher J.

2013-07-31

This report details the chemical and radionuclide contaminant results for the characterization of the 2013 Second Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time. Information from this characterization will be used by Saltstone Facility Engineering (SFE) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System.

Waste Acceptance System Requirements document (WASRD)

International Nuclear Information System (INIS)

1993-01-01

This Waste Acceptance System Requirements document (WA-SRD) describes the functions to be performed and the technical requirements for a Waste Acceptance System for accepting spent nuclear fuel (SNF) and high-level radioactive waste (HLW) into the Civilian Radioactive Waste Management System (CRWMS). This revision of the WA-SRD addresses the requirements for the acceptance of HLW. This revision has been developed as a top priority document to permit DOE's Office of Environmental Restoration and Waste Management (EM) to commence waste qualification runs at the Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF) in a timely manner. Additionally, this revision of the WA-SRD includes the requirements from the Physical System Requirements -- Accept Waste document for the acceptance of SNF. A subsequent revision will fully address requirements relative to the acceptance of SNF

Radiological performance assessment for the Z-Area Saltstone Disposal Facility

Energy Technology Data Exchange (ETDEWEB)

Cook, J.R.; Fowler, J.R. [Westinghouse Savannah River Co., Aiken, SC (United States)

1992-12-18

This radiological performance assessment (RPA) for the Savannah River Site (SRS) Saltstone Disposal Facility (SDF) was prepared in accordance with the requirements of Chapter III of the US Department of Energy Order 5820.2A. The Order specifies that an RPA should provide reasonable assurance that a low-level waste (LLW) disposal facility will comply with the performance objectives of the Order. The performance objectives require that: (1) exposures of the general public to radioactivity in the waste or released from the waste will not result in an effective dose equivalent of 25 mrem per year; (2) releases to the atmosphere will meet the requirements of 40 CFR 61; (3) inadvertent intruders will not be committed to an excess of an effective dose equivalent of 100 mrem per year from chronic exposure, or 500 mrem from a single acute exposure; and (4) groundwater resources will be protected in accordance with Federal, State and local requirements.

Radiological performance assessment for the Z-Area Saltstone Disposal Facility

International Nuclear Information System (INIS)

Cook, J.R.; Fowler, J.R.

1992-01-01

This radiological performance assessment (RPA) for the Savannah River Site (SRS) Saltstone Disposal Facility (SDF) was prepared in accordance with the requirements of Chapter III of the US Department of Energy Order 5820.2A. The Order specifies that an RPA should provide reasonable assurance that a low-level waste (LLW) disposal facility will comply with the performance objectives of the Order. The performance objectives require that: (1) exposures of the general public to radioactivity in the waste or released from the waste will not result in an effective dose equivalent of 25 mrem per year; (2) releases to the atmosphere will meet the requirements of 40 CFR 61; (3) inadvertent intruders will not be committed to an excess of an effective dose equivalent of 100 mrem per year from chronic exposure, or 500 mrem from a single acute exposure; and (4) groundwater resources will be protected in accordance with Federal, State and local requirements

Waste acceptance product specifications for vitrified high-level waste forms

International Nuclear Information System (INIS)

Applewhite-Ramsey, A.; Sproull, J.F.

1993-01-01

The Nuclear Waste Policy Act of 1982 mandated that all high-level waste (HLW) be sent to a federal geologic repository for permanent disposal. DOE published the Environmental Assessment in 1982 which identified borosilicate glass as the chosen HLW form. 1 In 1985 the Department of Energy instituted a Waste Acceptance Process to assure that DWPF glass waste forms would be acceptable to such a repository. This assurance was important since production of waste forms will precede repository construction and licensing. As part of this Waste Acceptance Process, the DOE Office of Civilian Radioactive Waste Management (RW) formed the Waste Acceptance Committee (WAC). The WAC included representatives from the candidate repository sites, the waste producing sites and DOE. The WAC was responsible for developing the Waste Acceptance Preliminary Specifications (WAPS) which defined the requirements the waste forms must meet to be compatible with the candidate repository geologies

Defense waste salt disposal at the Savannah River Plant

International Nuclear Information System (INIS)

Langton, C.A.; Dukes, M.D.

1984-01-01

A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. The disposal process includes emplacing the saltstone in engineered trenches above the water table but below grade at SRP. Design of the waste form and disposal system limits the concentration of salts and radionuclides in the groundwater so that EPA drinking water standards will not be exceeded at the perimeter of the disposal site. 10 references, 4 figures, 3 tables

Addendum to the Composite Analysis for the E-Area Vaults and Saltstone Disposal Facilities

International Nuclear Information System (INIS)

Cook, J.R.

2002-01-01

Revision 1 of the Composite Analysis (CA) Addendum has been prepared to respond to the U.S. Department of Energy (DOE) Low-Level Waste Disposal Facilities Federal Review Group review of the CA. This addendum to the composite analysis responds to the conditions of approval. The composite analysis was performed on the two active SRS low-level radioactive waste disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults Disposal Facility. The analysis calculated potential releases to the environment from all sources of residual radioactive material expected to remain in the General Separations Area (GSA). The GSA is the central part of the Savannah River Site and contains all of the waste disposal facilities, the chemical separation facilities and associated high-level waste storage facilities, as well as numerous other sources of radioactive material

Identification of permit and waste acceptance criteria provisions requiring modification for acceptance of commercial mixed waste

International Nuclear Information System (INIS)

1994-03-01

In October 1990, representatives of States and compact regions requested that the US Department of Energy (DOE) explore an agreement with host States and compact regions under which DOE would accept commercial mixed low-level radioactive waste (LLW) at DOE's own treatment and disposal facilities. A program for DOE management of commercial mixed waste is made potentially more attractive in light of the low commercial mixed waste volumes, high regulatory burdens, public opposition to new disposal sites, and relatively high cost of constructing commercial disposal facilities. Several studies were identified as essential in determining the feasibility of DOE accepting commercial mixed waste for disposal. The purpose of this report is to identify any current or proposed waste acceptance criteria (WAC) or Resource Conservation and Recovery Act (RCRA) provisions that would have to be modified for commercial mixed waste acceptance at specified DOE facilities. Following the introduction, Section 2 of this report (a) provides a background summary of existing and proposed mixed waste disposal facilities at each DOE site, and (b) summarizes the status of any RCRA Part B permit and WAC provisions relating to the disposal of mixed waste, including provisions relating to acceptance of offsite waste. Section 3 provides overall conclusions regarding the current status and permit modifications that must be implemented in order to grant DOE sites authority under their permits to accept commercial mixed waste for disposal. Section 4 contains a list of references

Results for the First, Second, and Third Quarter Calendar Year 2015 Tank 50H WAC slurry samples chemical and radionuclide contaminants

Energy Technology Data Exchange (ETDEWEB)

Crawford, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-02-18

This report details the chemical and radionuclide contaminant results for the characterization of the Calendar Year (CY) 2015 First, Second, and Third Quarter sampling of Tank 50H for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time. Information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering (D&S-FE) to support the transfer of low-level aqueous waste from Tank 50H to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50H Waste Characterization System. Previous memoranda documenting the WAC analyses results have been issued for these three samples.

Hanford Site solid waste acceptance criteria

International Nuclear Information System (INIS)

Ellefson, M.D.

1998-01-01

Order 5820.2A requires that each treatment, storage, and/or disposal facility (referred to in this document as TSD unit) that manages low-level or transuranic waste (including mixed waste and TSCA PCB waste) maintain waste acceptance criteria. These criteria must address the various requirements to operate the TSD unit in compliance with applicable safety and environmental requirements. This document sets forth the baseline criteria for acceptance of radioactive waste at TSD units operated by WMH. The criteria for each TSD unit have been established to ensure that waste accepted can be managed in a manner that is within the operating requirements of the unit, including environmental regulations, DOE Orders, permits, technical safety requirements, waste analysis plans, performance assessments, and other applicable requirements. Acceptance criteria apply to the following TSD units: the Low-Level Burial Grounds (LLBG) including both the nonregulated portions of the LLBG and trenches 31 and 34 of the 218-W-5 Burial Ground for mixed waste disposal; Central Waste Complex (CWC); Waste Receiving and Processing Facility (WRAP); and T Plant Complex. Waste from all generators, both from the Hanford Site and from offsite facilities, must comply with these criteria. Exceptions can be granted as provided in Section 1.6. Specific waste streams could have additional requirements based on the 1901 identified TSD pathway. These requirements are communicated in the Waste Specification Records (WSRds). The Hanford Site manages nonradioactive waste through direct shipments to offsite contractors. The waste acceptance requirements of the offsite TSD facility must be met for these nonradioactive wastes. This document does not address the acceptance requirements of these offsite facilities

NEVADA TEST SITE WASTE ACCEPTANCE CRITERIA

International Nuclear Information System (INIS)

U.S. DEPARTMENT OF ENERGY, NATIONAL NUCLEAR SECURITY ADMINISTRATION, NEVADA SITE OFFICE

2005-01-01

This document establishes the U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive and mixed waste for disposal. Mixed waste generated within the State of Nevada by NNSA/NSO activities is accepted for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the Nevada Test Site Area 3 and Area 5 Radioactive Waste Management Site for storage or disposal

Waste acceptance and logistics

International Nuclear Information System (INIS)

Carlson, James H.

1992-01-01

There are three major components which are normally highlighted when the Civilian Radioactive Waste Management Program is discussed - the repository, the monitored retrievable storage facility, and the transportation system. These are clearly the major physical system elements and they receive the greatest external attention. However, there will not be a successful, operative waste management system without fully operational waste acceptance plans and logistics arrangements. This paper will discuss the importance of developing, on a parallel basis to the normally considered waste management system elements, the waste acceptance and logistics arrangements to enable the timely transfer of spent nuclear fuel from more than one hundred and twenty waste generators to the Federal government. The paper will also describe the specific activities the Program has underway to make the necessary arrangements. (author)

Literature review of the potential impact of glycolic acid on the technetium chemistry of srs tank waste

International Nuclear Information System (INIS)

Nash, Charles A.; McCabe, Daniel J.

2017-01-01

This document presents a literature study of the impact of glycolate on technetium chemistry in the Savannah River Site (SRS) waste system and specifically Saltstone. A predominant portion of the Tc at SRS will be sent to the Saltstone Facility where it will be immobilized. The Tc in the tank waste is in the highly soluble chemical form of pertechnetate ion (TcO 4 - ) which is reduced by blast furnace slag (BFS) in Saltstone, rendering it highly insoluble and resistant to leaching.

Secondary Waste Cementitious Waste Form Data Package for the Integrated Disposal Facility Performance Assessment

Energy Technology Data Exchange (ETDEWEB)

Cantrell, Kirk J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Serne, R Jeffrey [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cozzi, Alex D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-05-16

A review of the most up-to-date and relevant data currently available was conducted to develop a set of recommended values for use in the Integrated Disposal Facility (IDF) performance assessment (PA) to model contaminant release from a cementitious waste form for aqueous wastes treated at the Hanford Effluent Treatment Facility (ETF). This data package relies primarily upon recent data collected on Cast Stone formulations fabricated with simulants of low-activity waste (LAW) and liquid secondary wastes expected to be produced at Hanford. These data were supplemented, when necessary, with data developed for saltstone (a similar grout waste form used at the Savannah River Site). Work is currently underway to collect data on cementitious waste forms that are similar to Cast Stone and saltstone but are tailored to the characteristics of ETF-treated liquid secondary wastes. Recommended values for key parameters to conduct PA modeling of contaminant release from ETF-treated liquid waste are provided.

Literature review of the potential impact of glycolic acid on the technetium chemistry of srs tank waste

Energy Technology Data Exchange (ETDEWEB)

Nash, Charles A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-10-09

This document presents a literature study of the impact of glycolate on technetium chemistry in the Savannah River Site (SRS) waste system and specifically Saltstone. A predominant portion of the Tc at SRS will be sent to the Saltstone Facility where it will be immobilized. The Tc in the tank waste is in the highly soluble chemical form of pertechnetate ion (TcO₄ ^-) which is reduced by blast furnace slag (BFS) in Saltstone, rendering it highly insoluble and resistant to leaching.

Waste acceptance product specifications for vitrified high-level waste forms

International Nuclear Information System (INIS)

Applewhite-Ramsey, A.; Sproull, J.F.

1994-01-01

The Department of Energy (DOE) Office of Environmental Restoration and Waste Management (EM) has developed Waste Acceptance Product Specifications (EM-WAPS). The EM-WAPS will be the basis for defining product acceptance criteria compatible with the requirements of the Civilian Radioactive Waste Management System (CRWMS). The relationship between the EM-WAPS and the CRWMS Systems Requirements document (WA-SRD) will be discussed. The impact of the EM-WAPS on the Savannah River Sit (SRS) Defense Waste Processing Facility's (DWPF) Waste Acceptance Program, Waste Qualification Run planning, and startup schedule will also be reported. 14 refs., 2 tabs

Nevada National Security Site Waste Acceptance Criteria

International Nuclear Information System (INIS)

2012-01-01

This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO), Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept DOE non-radioactive classified waste, DOE non-radioactive hazardous classified waste, DOE low-level radioactive waste (LLW), DOE mixed low-level waste (MLLW), and U.S. Department of Defense (DOD) classified waste for permanent disposal. Classified waste is the only waste accepted for disposal that may be non-radioactive and will be required to meet the waste acceptance criteria for radioactive waste as specified in this document. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project (WMP) at (702) 295-7063, and your call will be directed to the appropriate contact.

Nevada National Security Site Waste Acceptance Criteria

Energy Technology Data Exchange (ETDEWEB)

NSTec Environmental Management

2012-02-28

This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO), Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept DOE non-radioactive classified waste, DOE non-radioactive hazardous classified waste, DOE low-level radioactive waste (LLW), DOE mixed low-level waste (MLLW), and U.S. Department of Defense (DOD) classified waste for permanent disposal. Classified waste is the only waste accepted for disposal that may be non-radioactive and will be required to meet the waste acceptance criteria for radioactive waste as specified in this document. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project (WMP) at (702) 295-7063, and your call will be directed to the appropriate contact.

Method Evaluation And Field Sample Measurements For The Rate Of Movement Of The Oxidation Front In Saltstone

Energy Technology Data Exchange (ETDEWEB)

Almond, P. M. [Savannah River Site (SRS), Aiken, SC (United States); Kaplan, D. I. [Savannah River Site (SRS), Aiken, SC (United States); Langton, C. A. [Savannah River Site (SRS), Aiken, SC (United States); Stefanko, D. B. [Savannah River Site (SRS), Aiken, SC (United States); Spencer, W. A. [Savannah River Site (SRS), Aiken, SC (United States); Hatfield, A. [Clemson University, Clemson, SC (United States); Arai, Y. [Clemson University, Clemson, SC (United States)

2012-08-23

The objective of this work was to develop and evaluate a series of methods and validate their capability to measure differences in oxidized versus reduced saltstone. Validated methods were then applied to samples cured under field conditions to simulate Performance Assessment (PA) needs for the Saltstone Disposal Facility (SDF). Four analytical approaches were evaluated using laboratory-cured saltstone samples. These methods were X-ray absorption spectroscopy (XAS), diffuse reflectance spectroscopy (DRS), chemical redox indicators, and thin-section leaching methods. XAS and thin-section leaching methods were validated as viable methods for studying oxidation movement in saltstone. Each method used samples that were spiked with chromium (Cr) as a tracer for oxidation of the saltstone. The two methods were subsequently applied to field-cured samples containing chromium to characterize the oxidation state of chromium as a function of distance from the exposed air/cementitious material surface.

Verification of Sulfate Attack Penetration Rates for Saltstone Disposal Unit Modeling

Energy Technology Data Exchange (ETDEWEB)

Flach, G. P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-05-12

Recent Special Analysis modeling of Saltstone Disposal Units consider sulfate attack on concrete and utilize degradation rates estimated from Cementitious Barriers Partnership software simulations. This study provides an independent verification of those simulation results using an alternative analysis method and an independent characterization data source. The sulfate penetration depths estimated herein are similar to the best-estimate values in SRNL-STI-2013-00118 Rev. 2 and well below the nominal values subsequently used to define Saltstone Special Analysis base cases.

Hanford Site Solid Waste Acceptance Criteria

Energy Technology Data Exchange (ETDEWEB)

1993-11-17

This manual defines the Hanford Site radioactive, hazardous, and sanitary solid waste acceptance criteria. Criteria in the manual represent a guide for meeting state and federal regulations; DOE Orders; Hanford Site requirements; and other rules, regulations, guidelines, and standards as they apply to acceptance of radioactive and hazardous solid waste at the Hanford Site. It is not the intent of this manual to be all inclusive of the regulations; rather, it is intended that the manual provide the waste generator with only the requirements that waste must meet in order to be accepted at Hanford Site TSD facilities.

Hanford Site Solid Waste Acceptance Criteria

International Nuclear Information System (INIS)

1993-01-01

This manual defines the Hanford Site radioactive, hazardous, and sanitary solid waste acceptance criteria. Criteria in the manual represent a guide for meeting state and federal regulations; DOE Orders; Hanford Site requirements; and other rules, regulations, guidelines, and standards as they apply to acceptance of radioactive and hazardous solid waste at the Hanford Site. It is not the intent of this manual to be all inclusive of the regulations; rather, it is intended that the manual provide the waste generator with only the requirements that waste must meet in order to be accepted at Hanford Site TSD facilities

Large-scale demonstration of disposal of decontaminated salt as saltstone. Part I. Construction, loading, and capping of lysimeters

International Nuclear Information System (INIS)

Wolf, H.C.

1984-06-01

The installation phase of a large-scale demonstration of the disposal concept for decontaminated, low-level radioactive salt waste at the Savannah River Plant was completed in December 1983 and January 1984. The installation entailed immobilizing 7500 gallons of decontaminated salt solution with a blended cement formulation and pouring the resulting grout, saltstone, into three specially designed lysimeters for extended in-field leaching tests under natural conditions. 4 references, 35 figures, 4 tables

Nevada Test Site Waste Acceptance Criteria

International Nuclear Information System (INIS)

U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office

2005-01-01

This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive (LLW) and mixed waste (MW) for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NTS Area 3 and Area 5 Radioactive Waste Management Complex (RWMC) for storage or disposal

Nevada Test Site Waste Acceptance Criteria, December 2000

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-12-01

This document establishes the US Department of Energy, Nevada Operations Office waste acceptance criteria. The waste acceptance criteria provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive waste and mixed waste for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the Nevada Test Site Area 3 and Area 5 Radioactive Waste Management Sites for storage or disposal.

Nevada Test Site Waste Acceptance Criteria, December 2000

International Nuclear Information System (INIS)

2000-01-01

This document establishes the US Department of Energy, Nevada Operations Office waste acceptance criteria. The waste acceptance criteria provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive waste and mixed waste for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the Nevada Test Site Area 3 and Area 5 Radioactive Waste Management Sites for storage or disposal

Lessons Learned from an External Review of the Savannah River Site Saltstone Performance Assessment Program

International Nuclear Information System (INIS)

Cook, J.R.

2006-01-01

The Savannah River National Laboratory is actively working on a total revision of the Saltstone Performance Assessment. 'Lessons Learned' from the review are being applied to this effort. Examples of the areas in which significant new work is being done are development of a methodology to do probabilistic uncertainty analyses, employing quantitative analytical tools to represent long-term chemical degradation of both concrete and the Saltstone wasteform, and then using those tools to come to a better understanding of how changes in the vault and Saltstone will affect the performance of the overall disposal system over long periods of time. (authors)

Nevada National Security Site Waste Acceptance Criteria

Energy Technology Data Exchange (ETDEWEB)

none,

2013-06-01

This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO), Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept the following: • DOE hazardous and non-hazardous non-radioactive classified waste • DOE low-level radioactive waste (LLW) • DOE mixed low-level waste (MLLW) • U.S. Department of Defense (DOD) classified waste The LLW and MLLW listed above may also be classified waste. Classified waste is the only waste accepted for disposal that may be non-radioactive and shall be required to meet the waste acceptance criteria for radioactive waste as specified in this document. Classified waste may be sent to the NNSS as classified matter. Section 3.1.18 provides the requirements that must be met for permanent burial of classified matter. The NNSA/NFO and support contractors are available to assist the generator in understanding or interpreting this document. For assistance, please call the NNSA/NFO Environmental Management Operations (EMO) at (702) 295-7063, and the call will be directed to the appropriate contact.

Nevada National Security Site Waste Acceptance Criteria

International Nuclear Information System (INIS)

2013-01-01

This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO), Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept the following: DOE hazardous and non-hazardous non-radioactive classified waste; DOE low-level radioactive waste (LLW); DOE mixed low-level waste (MLLW); and, U.S. Department of Defense (DOD) classified waste. The LLW and MLLW listed above may also be classified waste. Classified waste is the only waste accepted for disposal that may be non-radioactive and shall be required to meet the waste acceptance criteria for radioactive waste as specified in this document. Classified waste may be sent to the NNSS as classified matter. Section 3.1.18 provides the requirements that must be met for permanent burial of classified matter. The NNSA/NFO and support contractors are available to assist the generator in understanding or interpreting this document. For assistance, please call the NNSA/NFO Environmental Management Operations (EMO) at (702) 295-7063, and the call will be directed to the appropriate contact.

Waste transmutation and public acceptance

International Nuclear Information System (INIS)

Pigford, T.H.

1991-01-01

The concept of transmuting radioactive wastes with reactors or accelerators is appealing. It has the potential of simplifying or eliminating problems of disposing of nuclear waste. The transmutation concept has been renewed vigorously at a time when national projects to dispose of high-level and transuranic waste are seriously delayed. In this period of tightening federal funds and program curtailments, skilled technical staffs are available at US Department of Energy (DOE) national laboratories and contractors to work on waste transmutation. If the claims of transmutation can be shown to be realistic, economically feasible, and capable of being implemented within the US institutional infrastructure, public acceptance of nuclear waste disposal may be enhanced. If the claims for transmutation are not substantiated, however, there will result a serious loss of credibility and an unjust exacerbation of public concerns about nuclear waste. The paper discusses the following topics: how public acceptance is achieved; the technical community and waste disposal; transmutation and technical communication; transmutation issues; technical fixes and public perception

Idaho CERCLA Disposal Facility Complex Waste Acceptance Criteria

Energy Technology Data Exchange (ETDEWEB)

W. Mahlon Heileson

2006-10-01

The Idaho Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Disposal Facility (ICDF) has been designed to accept CERCLA waste generated within the Idaho National Laboratory. Hazardous, mixed, low-level, and Toxic Substance Control Act waste will be accepted for disposal at the ICDF. The purpose of this document is to provide criteria for the quantities of radioactive and/or hazardous constituents allowable in waste streams designated for disposal at ICDF. This ICDF Complex Waste Acceptance Criteria is divided into four section: (1) ICDF Complex; (2) Landfill; (3) Evaporation Pond: and (4) Staging, Storage, Sizing, and Treatment Facility (SSSTF). The ICDF Complex section contains the compliance details, which are the same for all areas of the ICDF. Corresponding sections contain details specific to the landfill, evaporation pond, and the SSSTF. This document specifies chemical and radiological constituent acceptance criteria for waste that will be disposed of at ICDF. Compliance with the requirements of this document ensures protection of human health and the environment, including the Snake River Plain Aquifer. Waste placed in the ICDF landfill and evaporation pond must not cause groundwater in the Snake River Plain Aquifer to exceed maximum contaminant levels, a hazard index of 1, or 10-4 cumulative risk levels. The defined waste acceptance criteria concentrations are compared to the design inventory concentrations. The purpose of this comparison is to show that there is an acceptable uncertainty margin based on the actual constituent concentrations anticipated for disposal at the ICDF. Implementation of this Waste Acceptance Criteria document will ensure compliance with the Final Report of Decision for the Idaho Nuclear Technology and Engineering Center, Operable Unit 3-13. For waste to be received, it must meet the waste acceptance criteria for the specific disposal/treatment unit (on-Site or off-Site) for which it is destined.

Waste-acceptance criteria for greater confinement disposal

International Nuclear Information System (INIS)

Gilbert, T.L.; Meshkov, N.K.

1987-01-01

A methodology for establishing waste-acceptance criteria based on quantitative performance factors that characterize the confinement capabilities of a waste disposal site and facility has been developed. The methodology starts from the basic objective of protecting public health and safety by providing assurance that disposal of the waste will not result in a radiation dose to any member of the general public, in either the short or long term, in excess of an established basic dose limit. The method is based on an explicit, straight-forward, and quantitative relationship among individual risk, confinement capabilities, and waste characteristics. A key aspect of the methodology is introduction of a confinement factor that characterizes the overall confinement capability of a particular facility and can be used for quantitative assessments of the performance of different disposal sites and facilities, as well as for establishing site-specific waste acceptance criteria. Confinement factors are derived by means of site-specific pathway analyses. They make possible a direct and simple conversion of a basic dose limit into waste-acceptance criteria, specified as concentration limits on radionuclides in the waste streams and expressed in quantitative form as a function of parameters that characterize the site, facility design, waste containers, and waste form. Waste acceptance criteria can be represented visually as activity/time plots for various waste streams. These plots show the concentrations of radionuclides in a waste stream as a function of time and permit a visual, quantitative assessment of long-term performance, relative risks from different radionuclides in the waste stream, and contributions from ingrowth. 13 references, 7 figures

Waste-acceptance criteria for greater-confinement disposal

International Nuclear Information System (INIS)

Gilbert, T.L.; Meshkov, N.K.

1986-01-01

A methodology for establishing waste-acceptance criteria based on quantitative performance factors that characterize the confinement capabilities of a waste-disposal site and facility has been developed. The methodology starts from the basic objective of protecting public health and safety by providing assurance that dispsoal of the waste will not result in a radiation dose to any member of the general public, in either the short or long term, in excess of an established basic dose limit. The method is based on an explicit, straightforward, and quantitative relationship among individual risk, confinement capabilities, and waste characteristics. A key aspect of the methodology is the introduction of a confinement factor that characterizes the overall confinement capability of a particular facility and can be used for quantitative assessments of the performance of different disposal sites and facilities, as well as for establishing site-specific waste-acceptance criteria. Confinement factors are derived by means of site-specific pathway analyses. They make possible a direct and simple conversion of a basic dose limit into waste-acceptance criteria, specified as concentration limits on radionuclides in the waste streams and expressed in quantitative form as a function of parameters that characterize the site, facility design, waste containers, and waste form. Waste-acceptance criteria can be represented visually as activity/time plots for various waste streams. These plots show the concentrations of radionuclides in a waste stream as a function of time and permit a visual, quantitative assessment of long-term performance, relative risks from different radionuclides in the waste stream, and contributions from ingrowth. 13 refs

Estimated release from the saltstone landfill effect of landfill caps and landfill-cap/monolith-liner combinations

International Nuclear Information System (INIS)

Wilhite, E.L.

1985-01-01

The effect of capping the entire saltstone landfill is dependent on the effectiveness of the clay cap in preventing infiltration. A cap that is 99% effective will reduce releases from the saltstone landfill by a factor of 7.7. Several combinations of landfill design alterations will result in meeting ground water standards

Waste-acceptance criteria and risk-based thinking for radioactive-waste classification

International Nuclear Information System (INIS)

Lowenthal, M.D.

1998-01-01

The US system of radioactive-waste classification and its development provide a reference point for the discussion of risk-based thinking in waste classification. The official US system is described and waste-acceptance criteria for disposal sites are introduced because they constitute a form of de facto waste classification. Risk-based classification is explored and it is found that a truly risk-based system is context-dependent: risk depends not only on the waste-management activity but, for some activities such as disposal, it depends on the specific physical context. Some of the elements of the official US system incorporate risk-based thinking, but like many proposed alternative schemes, the physical context of disposal is ignored. The waste-acceptance criteria for disposal sites do account for this context dependence and could be used as a risk-based classification scheme for disposal. While different classes would be necessary for different management activities, the waste-acceptance criteria would obviate the need for the current system and could better match wastes to disposal environments saving money or improving safety or both

Results for the Fourth Quarter Calendar Year 2015 Tank 50H Salt Solution Sample

Energy Technology Data Exchange (ETDEWEB)

Crawford, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-01-11

In this memorandum, the chemical and radionuclide contaminant results from the Fourth Quarter Calendar Year 2015 (CY15) sample of Tank 50H salt solution are presented in tabulated form. The Fourth Quarter CY15 Tank 50H samples were obtained on October 29, 2015 and received at Savannah River National Laboratory (SRNL) on October 30, 2015. The information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering for the transfer of aqueous waste from Tank 50H to the Salt Feed Tank in the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the Task Technical and Quality Assurance Plan (TTQAP) for the Tank 50H saltstone task. The chemical and radionuclide contaminant results from the characterization of the Fourth Quarter Calendar Year 2015 (CY15) sampling of Tank 50H were requested by SRR personnel and details of the testing are presented in the SRNL Task Technical and Quality Assurance Plan.

Radwaste characteristics and Disposal Facility Waste Acceptance Criteria

International Nuclear Information System (INIS)

Sung, Suk Hyun; Jeong, Yi Yeong; Kim, Ki Hong

2008-01-01

The purpose of Radioactive Waste Acceptance Criteria (WAC) is to verify a radioactive waste compliance with radioactive disposal facility requirements in order to maintain a disposal facility's performance objectives and to ensure its safety. To develop WAC which is conformable with domestic disposal site conditions, we furthermore analysed the WAC of foreign disposal sites similar to the Kyung-Ju disposal site and the characteristics of various wastes which are being generated from Korea nuclear facilities. Radioactive WAC was developed in the technical cooperation with the Korea Atomic Energy Research Institute in consideration of characteristics of the wastes which are being generated from various facilities, waste generators' opinions and other conditions. The established criteria was also discussed and verified at an advisory committee which was comprised of some experts from universities, institutes and the industry. So radioactive WAC was developed to accept all wastes which are being generated from various nuclear facilities as much as possible, ensuring the safety of a disposal facility. But this developed waste acceptance criteria is not a criteria to accept all the present wastes generated from various nuclear facilities, so waste generators must seek an alternative treatment method for wastes which were not worth disposing of, and then they must treat the wastes more to be acceptable at a disposal site. The radioactive disposal facility WAC will continuously complement certain criteria related to a disposal concentration limit for individual radionuclide in order to ensure a long-term safety.

Hanford Site solid waste acceptance criteria

International Nuclear Information System (INIS)

Willis, N.P.; Triner, G.C.

1991-09-01

Westinghouse Hanford Company manages the Hanford Site solid waste treatment, storage, and disposal facilities for the US Department of Energy Field Office, Richland under contract DE-AC06-87RL10930. These facilities include radioactive solid waste disposal sites, radioactive solid waste storage areas and hazardous waste treatment, storage, and/or disposal facilities. This manual defines the criteria that must be met by waste generators for solid waste to be accepted by Westinghouse Hanford Company for treatment, storage and/or disposal facilities. It is to be used by all waste generators preparing radioactive solid waste for storage or disposal at the Hanford Site facilities and for all Hanford Site generators of hazardous waste. This manual is also intended for use by Westinghouse Hanford Company solid waste technical staff involved with approval and acceptance of solid waste. The criteria in this manual represent a compilation of state and federal regulations; US Department of Energy orders; Hanford Site requirements; and other rules, regulations, guidelines, and standards as they apply to management of solid waste. Where appropriate, these requirements are included in the manual by reference. It is the intent of this manual to provide guidance to the waste generator in meeting the applicable requirements

Preliminary waste acceptance requirements - Konrad repository project

International Nuclear Information System (INIS)

Brennecke, P.W.; Warnecke, E.H.

1991-01-01

In Germany, the planned Konrad repository is proposed for the disposal of all types of radioactive wastes whose thermal influence upon the host rock is negligible. The Bundesamt fuer Strahlenschutz has established Preliminary Waste Acceptance Requirements (as of April 1990) for this facility. The respective requirements were developed on the basis of the results of site-specific safety assessments. They include general requirements on the waste packages to be disposed of as well as more specific requirements on the waste forms, the packaging and the radionuclide inventory per waste package. In addition, the delivery of waste packages was regulated. An outline of the structure and the elements of the Preliminary Waste Acceptance Requirements of April 1990 is given including comments on their legal status. (Author)

Nevada National Security Site Waste Acceptance Criteria

Energy Technology Data Exchange (ETDEWEB)

NSTec Environmental Management

2010-09-03

This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NNSSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NNSS Area 3 and Area 5 Radioactive Waste Management Complex for disposal. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project at (702) 295-7063 or fax to (702) 295-1153.

Nevada National Security Site Waste Acceptance Criteria

International Nuclear Information System (INIS)

2010-01-01

This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NNSSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NNSS Area 3 and Area 5 Radioactive Waste Management Complex for disposal. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project at (702) 295-7063 or fax to (702) 295-1153.

Nevada National Security Site Waste Acceptance Criteria

International Nuclear Information System (INIS)

2011-01-01

This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NNSSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NNSS Area 3 and Area 5 Radioactive Waste Management Complex for disposal. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project at (702) 295-7063 or fax to (702) 295-1153.

Miscibility Evaluation Of The Next Generation Solvent With Polymers Currently Used At DWPF, MCU, And Saltstone

Energy Technology Data Exchange (ETDEWEB)

Fondeur, F. F.

2013-04-17

The Office of Waste Processing, within the Office of Technology Innovation and Development, funded the development of an enhanced Caustic-Side Solvent Extraction (CSSX) solvent for deployment at the Savannah River Site for removal of cesium from High Level Waste. This effort lead to the development of the Next Generation Solvent (NGS) with Tris (3,7-dimethyl octyl) guanidine (TiDG). The first deployment target for the NGS solvent is within the Modular CSSX Unit (MCU). Deployment of a new chemical within an existing facility requires verification that the new chemical components are compatible with the installed equipment. In the instance of a new organic solvent, the primary focus is on compatibility of the solvent with organic polymers used in the affected facility. This report provides the calculated data from exposing these polymers to the Next Generation Solvent. An assessment of the dimensional stability of polymers known to be used or present in the MCU, Defense Waste Processing Facility (DWPF), and Saltstone facilities that will be exposed to the NGS showed that TiDG could selectively affect the elastomers and some thermoplastics to varying extents, but the typical use of these polymers in a confined geometry will likely prevent the NGS from impacting component performance. The polymers identified as of primary concern include Grafoil® (flexible graphite), Tefzel®, Isolast®, ethylene-propylene-diene monomer (EPDM) rubber, nitrile-butadiene rubber (NBR), styrene-butadiene rubber (SBR), ultra high molecular weight polyethylene (UHMWPE), and fluorocarbon rubber (FKM). Certain polymers like NBR and EPDM were found to interact mildly with NGS but their calculated swelling and the confined geometry will impede interaction with NGS. In addition, it was found that Vellumoid (cellulose fibers-reinforced glycerin and protein) may leach protein and Polyvinyl Chloride (PVC) may leach plasticizer (such as Bis-Ethylhexyl-Phthalates) into the NGS solvent. Either case

Nevada Test Site Waste Acceptance Criteria (NTSWAC)

Energy Technology Data Exchange (ETDEWEB)

NNSA/NSO Waste Management Project

2008-06-01

This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, Nevada Test Site Waste Acceptance Criteria (NTSWAC). The NTSWAC provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive (LLW) and LLW Mixed Waste (MW) for disposal.

Nevada Test Site Waste Acceptance Criteria (NTSWAC)

International Nuclear Information System (INIS)

NNSA/NSO Waste Management Project

2008-01-01

This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, Nevada Test Site Waste Acceptance Criteria (NTSWAC). The NTSWAC provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive (LLW) and LLW Mixed Waste (MW) for disposal

Nevada test site waste acceptance criteria

International Nuclear Information System (INIS)

1996-01-01

This document provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive and mixed waste for disposal; and transuranic and transuranic mixed waste for interim storage at the NTS. Review each section of this document. This document is not intended to include all of the requirements; rather, it is meant as a guide toward meeting the regulations. All references in this document should be observed to avoid omission of requirements on which acceptance or rejection of waste will be based. The Department of Energy/Nevada Operations Office (DOE/NV) and support contractors are available to assist you in understanding or interpreting this document

Nevada test site waste acceptance criteria

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

This document provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive and mixed waste for disposal; and transuranic and transuranic mixed waste for interim storage at the NTS. Review each section of this document. This document is not intended to include all of the requirements; rather, it is meant as a guide toward meeting the regulations. All references in this document should be observed to avoid omission of requirements on which acceptance or rejection of waste will be based. The Department of Energy/Nevada Operations Office (DOE/NV) and support contractors are available to assist you in understanding or interpreting this document.

Nevada Test Site Waste Acceptance Criteria

International Nuclear Information System (INIS)

U.S. Department of Energy, Nevada Operations Office, Waste Acceptance Criteria

1999-01-01

This document provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive and mixed waste for disposal; and transuranic and transuranic mixed waste for interim storage at the Nevada Test Site

Acceptable knowledge document for INEEL stored transuranic waste - Rocky Flats Plant waste. Revision 2

International Nuclear Information System (INIS)

1998-01-01

This document and supporting documentation provide a consistent, defensible, and auditable record of acceptable knowledge for waste generated at the Rocky Flats Plant which is currently in the accessible storage inventory at the Idaho National Engineering and Environmental Laboratory. The inventory consists of transuranic (TRU) waste generated from 1972 through 1989. Regulations authorize waste generators and treatment, storage, and disposal facilities to use acceptable knowledge in appropriate circumstances to make hazardous waste determinations. Acceptable knowledge includes information relating to plant history, process operations, and waste management, in addition to waste-specific data generated prior to the effective date of the RCRA regulations. This document is organized to provide the reader a comprehensive presentation of the TRU waste inventory ranging from descriptions of the historical plant operations that generated and managed the waste to specific information about the composition of each waste group. Section 2 lists the requirements that dictate and direct TRU waste characterization and authorize the use of the acceptable knowledge approach. In addition to defining the TRU waste inventory, Section 3 summarizes the historical operations, waste management, characterization, and certification activities associated with the inventory. Sections 5.0 through 26.0 describe the waste groups in the inventory including waste generation, waste packaging, and waste characterization. This document includes an expanded discussion for each waste group of potential radionuclide contaminants, in addition to other physical properties and interferences that could potentially impact radioassay systems

Acceptable knowledge document for INEEL stored transuranic waste -- Rocky Flats Plant waste. Revision 2

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-23

This document and supporting documentation provide a consistent, defensible, and auditable record of acceptable knowledge for waste generated at the Rocky Flats Plant which is currently in the accessible storage inventory at the Idaho National Engineering and Environmental Laboratory. The inventory consists of transuranic (TRU) waste generated from 1972 through 1989. Regulations authorize waste generators and treatment, storage, and disposal facilities to use acceptable knowledge in appropriate circumstances to make hazardous waste determinations. Acceptable knowledge includes information relating to plant history, process operations, and waste management, in addition to waste-specific data generated prior to the effective date of the RCRA regulations. This document is organized to provide the reader a comprehensive presentation of the TRU waste inventory ranging from descriptions of the historical plant operations that generated and managed the waste to specific information about the composition of each waste group. Section 2 lists the requirements that dictate and direct TRU waste characterization and authorize the use of the acceptable knowledge approach. In addition to defining the TRU waste inventory, Section 3 summarizes the historical operations, waste management, characterization, and certification activities associated with the inventory. Sections 5.0 through 26.0 describe the waste groups in the inventory including waste generation, waste packaging, and waste characterization. This document includes an expanded discussion for each waste group of potential radionuclide contaminants, in addition to other physical properties and interferences that could potentially impact radioassay systems.

US Department of Energy acceptance of commercial transuranic waste

International Nuclear Information System (INIS)

Taboas, A.L.; Bennett, W.S.; Brown, C.M.

1980-02-01

Contaminated transuranic wastes generated as a result of non-defense activities have been disposed of by shallow land burial at a commercially operated (NECO) facility located on the Hanford federal reservation, which is licensed by the State of Washington and by the NRC. About 15,000 ft 3 of commercial TRU waste have been generated each year, but generation for the next three years could triple due to decontamination and decommissioning scheduled to start in 1980. Disposal at other commercial burial sites has been precluded due to sites closing or prohibitions on acceptance of transuranic wastes. The State of Washington recently modified the NECO-Hanford operating license, effective February 29, 1980, to provide that radioactive wastes contaminated with transuranics in excess of 10 nCi/g will not be accepted for disposal. Consistent with the state policy, the NRC amended the NECO special nuclear material license so that Pu in excess of 10n Ci/g cannot be accepted after February 29, 1980. As a result, NRC requested DOE to examine the feasibility of accepting these wastes at a DOE operated site. TRU wastes accepted by the DOE would be placed in retrievable storage in accordance with DOE policy which requires retrievable storage of transuranic wastes pending final disposition in a geologic repository. DOE transuranic wastes are stored at six major DOE sites: INEL, Hanford, LASL, NTS, ORNL, and SRP. A specific site for receiving commercial TRU waste has not yet been selected. Shipments to DOE-Hanford would cause the least disruption to past practices. Commercial TRU wastes would be subject to waste form and packaging criteria established by the DOE. The waste generators would be expected to incur all applicable costs for DOE to take ownership of the waste, and provide storage, processing, and repository disposal. The 1980 charge to generators for DOE acceptance of commercial TRU waste is $147 per cubic foot

EVALUATION OF SULFATE ATTACK ON SALTSTONE VAULT CONCRETE AND SALTSTONESIMCO TECHNOLOGIES, INC. PART1 FINAL REPORT

International Nuclear Information System (INIS)

Langton, C.

2008-01-01

This report summarizes the preliminary results of a durability analysis performed by SIMCO Technologies Inc. to assess the effects of contacting saltstone Vaults 1/4 and Disposal Unit 2 concretes with highly alkaline solutions containing high concentrations of dissolved sulfate. The STADIUM(reg s ign) code and data from two surrogate concretes which are similar to the Vaults 1/4 and Disposal Unit 2 concretes were used in the preliminary durability analysis. Simulation results for these surrogate concrete mixes are provided in this report. The STADIUM(reg s ign) code will be re-run using transport properties measured for the SRS Vaults 1/4 and Disposal Unit 2 concrete samples after SIMCO personnel complete characterization testing on samples of these materials. Simulation results which utilize properties measured for samples of Vaults 1/4 and Disposal Unit 2 concretes will be provided in Revision 1 of this report after property data become available. The modeling performed to date provided the following information on two concrete mixes that will be used to support the Saltstone PA: (1) Relationship between the rate of advancement of the sulfate front (depth of sulfate ion penetration into the concrete) and the rate of change of the concrete permeability and diffusivity. (2) Relationship between the sulfate ion concentration in the corrosive leachate and the rate of the sulfate front progression. (3) Equation describing the change in hydraulic properties (hydraulic conductivity and diffusivity) as a function of sulfate ion concentration in the corrosive leachate. These results have been incorporated into the current Saltstone PA analysis by G. Flach (Flach, 2008). In addition, samples of the Saltstone Vaults 1/4 and Disposal Unit 2 concretes have been prepared by SIMCO Technologies, Inc. Transport and physical properties for these materials are currently being measured and sulfate exposure testing to three high alkaline, high sulfate leachates provided by SRNL is

Atmospheric Pathway Screening Analysis for Saltstone Disposal Facility Vault 4

International Nuclear Information System (INIS)

COOK, JAMES

2004-01-01

A sequential screening process using a methodology developed by the National Council on Radiation Protection and Measurements, professional judgment and process knowledge has been used to produce a list of radionuclides requiring detailed analysis to derive disposal limits for the Saltstone Disposal Facility based on the atmospheric pathway

Preliminary waste acceptance requirements for the planned Konrad repository

International Nuclear Information System (INIS)

Warnecke, E.; Brennecke, P.

1987-01-01

The Physikalisch-Technische Bundesanstalt (PTB) has established Preliminary Waste Acceptance Requirements for the planned Konrad repository. These requirements were developed, in accordance with the Safety Criteria of the Reactor Safety Commission, with the help of a site specific safety assessment; they are under the reservation of the plan approval procedure, which is still in progress. In developing waste acceptance requirements, the PTB fulfills one of its duties as the institute responsible for waste disposal and gives guidelines for waste conditioning to waste producers and conditioners. (orig.) [de

Radioactive waste disposal and public acceptance aspects

Energy Technology Data Exchange (ETDEWEB)

Ulhoa, Barbara M.A.; Aleixo, Bruna L.; Mourao, Rogerio P.; Ferreira, Vinicius V.M., E-mail: mouraor@cdtn.b, E-mail: vvmf@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

Part of the public opinion around the world considers the wastes generated due to nuclear applications as the biggest environmental problem of the present time. The development of a solution that satisfies everybody is a great challenge, in that obtaining public acceptance for nuclear enterprises is much more challenging than solving the technical issues involved. Considering that the offering of a final solution that closes the radioactive waste cycle has a potentially positive impact on public opinion, the objective of this work is to evaluate the amount of the radioactive waste volume disposed in a five-year period in several countries and gauge the public opinion regarding nuclear energy. The results show that the volume of disposed radioactive waste increased, a fact that stresses the importance of promoting discussions about repositories and public acceptance. (author)

Radioactive waste disposal and public acceptance aspects

International Nuclear Information System (INIS)

Ulhoa, Barbara M.A.; Aleixo, Bruna L.; Mourao, Rogerio P.; Ferreira, Vinicius V.M.

2011-01-01

Part of the public opinion around the world considers the wastes generated due to nuclear applications as the biggest environmental problem of the present time. The development of a solution that satisfies everybody is a great challenge, in that obtaining public acceptance for nuclear enterprises is much more challenging than solving the technical issues involved. Considering that the offering of a final solution that closes the radioactive waste cycle has a potentially positive impact on public opinion, the objective of this work is to evaluate the amount of the radioactive waste volume disposed in a five-year period in several countries and gauge the public opinion regarding nuclear energy. The results show that the volume of disposed radioactive waste increased, a fact that stresses the importance of promoting discussions about repositories and public acceptance. (author)

Nevada Test Site Waste Acceptance Criteria (NTSWAC), Rev. 7-01

Energy Technology Data Exchange (ETDEWEB)

NSTec Environmental Management

2009-05-01

This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, Nevada Test Site Waste Acceptance Criteria (NTSWAC). The NTSWAC provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NTSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NTS Area 3 and Area 5 Radioactive Waste Management Complex for disposal.

Nevada Test Site Waste Acceptance Criteria (NTSWAC), Rev. 7-01

International Nuclear Information System (INIS)

2009-01-01

This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, Nevada Test Site Waste Acceptance Criteria (NTSWAC). The NTSWAC provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NTSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NTS Area 3 and Area 5 Radioactive Waste Management Complex for disposal.

Standardization of waste acceptance test methods by the Materials Characterization Center

International Nuclear Information System (INIS)

Slate, S.C.

1985-01-01

This paper describes the role of standardized test methods in demonstrating the acceptability of high-level waste (HLW) forms for disposal. Key waste acceptance tests are standardized by the Materials Characterization Center (MCC), which the US Department of Energy (DOE) has established as the central agency in the United States for the standardization of test methods for nuclear waste materials. This paper describes the basic three-step process that is used to show that waste is acceptable for disposal and discusses how standardized tests are used in this process. Several of the key test methods and their areas of application are described. Finally, future plans are discussed for using standardized tests to show waste acceptance. 9 refs., 1 tab

Nevada test site defense waste acceptance criteria, certification, and transfer requirements

International Nuclear Information System (INIS)

1988-10-01

The Nevada Test Site (NTS) Defense Waste Acceptance Criteria, Certification and Transfer Requirements establishes procedures and criteria for safe transfer, disposal, and storage of defense transuranic, low-level, and mixed waste at the NTS. Included are an overview of the NTS defense waste management program; the NTS waste acceptance criteria for transuranic, low-level, and mixed wastes; waste certification requirements and guidance; application to submit waste; and requirements for waste transfer and receipt. 5 figs., 16 tabs

Nitrate Diffusional Releases from the Saltstone Facility, Vault 2, with Respect to Different Concrete Wall Thicknesses

International Nuclear Information System (INIS)

ROBERT, HIERGESELL

2005-01-01

To assist the Saltstone Vault 2 Design Team, an investigation was conducted to evaluate the effectiveness of alternative concrete wall thicknesses in limiting nitrate diffusion away from the planned facility. While the current design calls for 18-inch concrete walls, alternative thicknesses of 12-in, 8-in, and 6-in were evaluated using a simplified 1-D numerical model. To serve as a guide for Saltstone Vault 2 conceptual design, the results of this investigation were applied to Saltstone Vault 4 to determine what the hypothetical limits would be for concrete wall thicknesses thinner than the planned 18-inches. This was accomplished by adjusting the Vault 4 Limits, based on the increased nitrate diffusion rates through the thinner concrete walls, such that the 100-m well limit of 44 mg/L of nitrate as nitrate was not exceeded. The implication of these preliminary results is that as thinner vault walls are implemented there is a larger release of nitrate, thus necessitating optimal vault placement to minimize the number of vaults placed along a single groundwater flow path leading to the discharge zone

Documentation of acceptable knowledge for LANL Plutonium Facility transuranic waste streams

International Nuclear Information System (INIS)

Montoya, A.J.; Gruetzmacher, K.; Foxx, C.; Rogers, P.S.Z.

1998-01-01

Characterization of transuranic waste from the LANL Plutonium Facility for certification and transportation to WIPP includes the use of acceptable knowledge as specified in the WIPP Quality Assurance Program Plan. In accordance with a site-specific procedure, documentation of acceptable knowledge for retrievably stored and currently generated transuranic waste streams is in progress at LANL. A summary overview of the transuranic waste inventory is complete and documented in the Sampling Plan. This document also includes projected waste generation, facility missions, waste generation processes, flow diagrams, times, and material inputs. The second part of acceptable knowledge documentation consists of assembling more detailed acceptable knowledge information into auditable records and is expected to require several years to complete. These records for each waste stream must support final assignment of waste matrix parameters, EPA hazardous waste numbers, and radionuclide characterization. They must also include a determination whether waste streams are defense waste streams for compliance with the WIPP Land Withdrawal Act. The LANL Plutonium Facility's mission is primarily plutonium processing in basic special nuclear material (SNM) research activities to support national defense and energy programs. It currently has about 100 processes ranging from SNM recovery from residues to development of plutonium 238 heat sources for space applications. Its challenge is to characterize and certify waste streams from such diverse and dynamic operations using acceptable knowledge. This paper reports the progress on the certification of the first of these waste streams to the WIPP WAC

Contact-Handled Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

2005-01-01

The purpose of this document is to summarize the waste acceptance criteria applicable to the transportation, storage, and disposal of contact-handled transuranic (CH-TRU) waste at the Waste Isolation Pilot Plant (WIPP). These criteria serve as the U.S. Department of Energy's (DOE) primary directive for ensuring that CH-TRU waste is managed and disposed of in a manner that protects human health and safety and the environment.The authorization basis of WIPP for the disposal of CH-TRU waste includes the U.S.Department of Energy National Security and Military Applications of Nuclear EnergyAuthorization Act of 1980 (reference 1) and the WIPP Land Withdrawal Act (LWA;reference 2). Included in this document are the requirements and associated criteriaimposed by these acts and the Resource Conservation and Recovery Act (RCRA,reference 3), as amended, on the CH-TRU waste destined for disposal at WIPP.|The DOE TRU waste sites must certify CH-TRU waste payload containers to thecontact-handled waste acceptance criteria (CH-WAC) identified in this document. Asshown in figure 1.0, the flow-down of applicable requirements to the CH-WAC istraceable to several higher-tier documents, including the WIPP operational safetyrequirements derived from the WIPP CH Documented Safety Analysis (CH-DSA;reference 4), the transportation requirements for CH-TRU wastes derived from theTransuranic Package Transporter-Model II (TRUPACT-II) and HalfPACT Certificates ofCompliance (references 5 and 5a), the WIPP LWA (reference 2), the WIPP HazardousWaste Facility Permit (reference 6), and the U.S. Environmental Protection Agency(EPA) Compliance Certification Decision and approval for PCB disposal (references 7,34, 35, 36, and 37). The solid arrows shown in figure 1.0 represent the flow-down of allapplicable payload container-based requirements. The two dotted arrows shown infigure 1.0 represent the flow-down of summary level requirements only; i.e., the sitesmust reference the regulatory source

Contact-Handled Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant

Energy Technology Data Exchange (ETDEWEB)

Washington TRU Solutions LLC

2005-12-29

The purpose of this document is to summarize the waste acceptance criteria applicable to the transportation, storage, and disposal of contact-handled transuranic (CH-TRU) waste at the Waste Isolation Pilot Plant (WIPP). These criteria serve as the U.S. Department of Energy's (DOE) primary directive for ensuring that CH-TRU waste is managed and disposed of in a manner that protects human health and safety and the environment.The authorization basis of WIPP for the disposal of CH-TRU waste includes the U.S.Department of Energy National Security and Military Applications of Nuclear EnergyAuthorization Act of 1980 (reference 1) and the WIPP Land Withdrawal Act (LWA;reference 2). Included in this document are the requirements and associated criteriaimposed by these acts and the Resource Conservation and Recovery Act (RCRA,reference 3), as amended, on the CH-TRU waste destined for disposal at WIPP.|The DOE TRU waste sites must certify CH-TRU waste payload containers to thecontact-handled waste acceptance criteria (CH-WAC) identified in this document. Asshown in figure 1.0, the flow-down of applicable requirements to the CH-WAC istraceable to several higher-tier documents, including the WIPP operational safetyrequirements derived from the WIPP CH Documented Safety Analysis (CH-DSA;reference 4), the transportation requirements for CH-TRU wastes derived from theTransuranic Package Transporter-Model II (TRUPACT-II) and HalfPACT Certificates ofCompliance (references 5 and 5a), the WIPP LWA (reference 2), the WIPP HazardousWaste Facility Permit (reference 6), and the U.S. Environmental Protection Agency(EPA) Compliance Certification Decision and approval for PCB disposal (references 7,34, 35, 36, and 37). The solid arrows shown in figure 1.0 represent the flow-down of allapplicable payload container-based requirements. The two dotted arrows shown infigure 1.0 represent the flow-down of summary level requirements only; i.e., the sitesmust reference the regulatory source

Multi-discipline Waste Acceptance Process at the Nevada National Security Site - 13573

Energy Technology Data Exchange (ETDEWEB)

Carilli, Jhon T. [US Department Of Energy, Nevada Site Office, P. O. Box 98518, Las Vegas, Nevada 89193-8518 (United States); Krenzien, Susan K. [Navarro-Intera, LLC, P. O. Box 98952, Las Vegas, Nevada 89193-8952 (United States)

2013-07-01

The Nevada National Security Site low-level radioactive waste disposal facility acceptance process requires multiple disciplines to ensure the protection of workers, the public, and the environment. These disciplines, which include waste acceptance, nuclear criticality, safety, permitting, operations, and performance assessment, combine into the overall waste acceptance process to assess low-level radioactive waste streams for disposal at the Area 5 Radioactive Waste Management Site. Four waste streams recently highlighted the integration of these disciplines: the Oak Ridge Radioisotope Thermoelectric Generators and Consolidated Edison Uranium Solidification Project material, West Valley Melter, and classified waste. (authors)

Results For The Third Quarter Calendar Year 2016 Tank 50H Salt Solution Sample

Energy Technology Data Exchange (ETDEWEB)

Crawford, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-10-13

In this memorandum, the chemical and radionuclide contaminant results from the Third Quarter Calendar Year 2016 (CY16) sample of Tank 50H salt solution are presented in tabulated form. The Third Quarter CY16 Tank 50H samples (a 200 mL sample obtained 6” below the surface (HTF-5-16-63) and a 1 L sample obtained 66” from the tank bottom (HTF-50-16-64)) were obtained on July 14, 2016 and received at Savannah River National Laboratory (SRNL) on the same day. Prior to obtaining the samples from Tank 50H, a single pump was run at least 4.4 hours, and the samples were pulled immediately after pump shut down. The information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering for the transfer of aqueous waste from Tank 50H to the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the Task Technical and Quality Assurance Plan (TTQAP) for the Tank 50H saltstone task. The chemical and radionuclide contaminant results from the characterization of the Third Quarter CY16 sampling of Tank 50H were requested by Savannah River Remediation (SRR) personnel and details of the testing are presented in the SRNL TTQAP.

SPEEDUP simulation of liquid waste batch processing. Revision 1

International Nuclear Information System (INIS)

Shannahan, K.L.; Aull, J.E.; Dimenna, R.A.

1994-01-01

The Savannah River Site (SRS) has accumulated radioactive hazardous waste for over 40 years during the time SRS made nuclear materials for the United States Department of Energy (DOE) and its predecessors. This waste is being stored as caustic slurry in a large number of 1 million gallon steel tanks, some of which were initially constructed in the early 1950's. SRS and DOE intend to clean up the Site and convert this waste into stable forms which then can be safely stored. The liquid waste will be separated into a partially decontaminated low-level and radioactive high-level waste in one feed preparation operation, In-Tank Precipitation. The low-level waste will be used to make a concrete product called saltstone in the Saltstone Facility, a part of the Defense Waste Processing Facility (DWPF). The concrete will be poured into large vaults, where it will be permanently stored. The high-level waste will be added to glass-formers and waste slurry solids from another feed preparation operation, Extended Sludge Processing. The mixture will then be converted to a stable borosilicate glass by a vitrification process that is the other major part of the DWPF. This glass will be poured into stainless steel canisters and sent to a temporary storage facility prior to delivery to a permanent underground storage site

Waste management facility acceptance - some findings

International Nuclear Information System (INIS)

Sigmon, B.

1987-01-01

Acceptance of waste management facilities remains a significant problem, despite years of efforts to reassure potential host communities. The tangible economic benefits from jobs, taxes, and expenditures are generally small, while the intangible risks of environmental or other impacts are difficult to evaluate and understand. No magic formula for winning local acceptance has yet been found. Limited case study and survey work does suggest some pitfalls to be avoided and some directions to be pursued. Among the most significant is the importance that communities place on controlling their own destiny. Finding a meaningful role for communities in the planning and operation of waste management facilities is a challenge that would-be developers should approach with the same creativity that characterizes their technical efforts

Attention: no radioactive waste accepted on 7 September

CERN Multimedia

2012-01-01

Anouncement by the RW section of the Radiation Protection Group: The Treatment Centre for Radioactive Waste will not be accepting waste on Friday, 7 September 2012. Thank you for adjusting your activities accordingly.

Nevada Test Site waste acceptance criteria [Revision 1

Energy Technology Data Exchange (ETDEWEB)

None

1997-08-01

Revision one updates the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive and mixed waste for disposal; and transuranic and transuranic mixed waste for interim storage at the NTS. Review each section of this document. This document is not intended to include all of the requirements; rather, it is meant as a guide toward meeting the regulations. All references in this document should be observed to avoid omission of requirements on which acceptance or rejection of waste will be based. The Department of Energy/Nevada Operations Office (DOE/NV) and support contractors are available to assist you in understanding or interpreting this document.

Nevada Test Site waste acceptance criteria [Revision 1

International Nuclear Information System (INIS)

None

1997-01-01

Revision one updates the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive and mixed waste for disposal; and transuranic and transuranic mixed waste for interim storage at the NTS. Review each section of this document. This document is not intended to include all of the requirements; rather, it is meant as a guide toward meeting the regulations. All references in this document should be observed to avoid omission of requirements on which acceptance or rejection of waste will be based. The Department of Energy/Nevada Operations Office (DOE/NV) and support contractors are available to assist you in understanding or interpreting this document

Anticipating Potential Waste Acceptance Criteria for Defense Spent Nuclear Fuel

International Nuclear Information System (INIS)

Rechard, R.P.; Lord, M.E.; Stockman, C.T.; McCurley, R.D.

1997-01-01

The Office of Environmental Management of the U.S. Department of Energy is responsible for the safe management and disposal of DOE owned defense spent nuclear fuel and high level waste (DSNF/DHLW). A desirable option, direct disposal of the waste in the potential repository at Yucca Mountain, depends on the final waste acceptance criteria, which will be set by DOE's Office of Civilian Radioactive Waste Management (OCRWM). However, evolving regulations make it difficult to determine what the final acceptance criteria will be. A method of anticipating waste acceptance criteria is to gain an understanding of the DOE owned waste types and their behavior in a disposal system through a performance assessment and contrast such behavior with characteristics of commercial spent fuel. Preliminary results from such an analysis indicate that releases of 99Tc and 237Np from commercial spent fuel exceed those of the DSNF/DHLW; thus, if commercial spent fuel can meet the waste acceptance criteria, then DSNF can also meet the criteria. In large part, these results are caused by the small percentage of total activity of the DSNF in the repository (1.5%) and regulatory mass (4%), and also because commercial fuel cladding was assumed to provide no protection

INTERNATIONAL PROGRAM: SUMMARY REPORT ON THE PROPERTIES OF CEMENTITIOUS WASTE FORMS

International Nuclear Information System (INIS)

Harbour, J

2007-01-01

This report provides a summary of the results on the properties of cementitious waste forms obtained as part of the International Program. In particular, this report focuses on the results of Task 4 of the Program that was initially entitled ''Improved Retention of Key Contaminants of Concern in Low Temperature Immobilized Waste Forms''. Task 4 was a joint program between Khlopin Radium Institute and the Savannah River National Laboratory. The task evolved during this period into a study of cementitious waste forms with an expanded scope that included heat of hydration and fate and transport modeling. This report provides the results for Task 4 of the International Program as of the end of FY06 at which time funding for Task 4 was discontinued due to the needs of higher priority tasks within the International Program. Consequently, some of the subtasks were only partially completed, but it was considered important to capture the results up to this point in time. Therefore, this report serves as the closeout report for Task 4. The degree of immobilization of Tc-99 within the Saltstone waste form was measured through monolithic and crushed grout leaching tests. An effective diffusion coefficient of 4.8 x 10 -12 (Leach Index of 11.4) was measured using the ANSI/ANS-16.1 protocol which is comparable with values obtained for tank closure grouts using a dilute salt solution. The leaching results show that, in the presence of concentrated salt solutions such as those that will be processed at the Saltstone Production Facility, blast furnace slag can effectively reduce pertechnetate to the immobile +4 oxidation state. Leaching tests were also initiated to determine the degree of immobilization of selenium in the Saltstone waste form. Results were obtained for the upper bound of projected selenium concentration (∼5 x 10 -3 M) in the salt solution that will be treated at Saltstone. The ANSI/ANS 16.1 leaching tests provided a value for the effective diffusivity of ∼5 x 10

Acceptance criteria for disposal of radioactive wastes in shallow ground and rock cavities

International Nuclear Information System (INIS)

1985-01-01

This document provides an overview of basic information related to waste acceptance criteria for disposal in shallow ground and rock cavity repositories, consisting of a discussion of acceptable waste types. The last item includes identification of those waste characteristics which may influence the performance of the disposal system and as such are areas of consideration for criteria development. The material is presented in a manner similar to a safety assessment. Waste acceptance criteria aimed at limiting the radiation exposure to acceptable levels are presented for each pathway. Radioactive wastes considered here are low-level radioactive wastes and intermediate-level radioactive wastes from nuclear fuel cycle operations and applications of radionuclides in research, medicine and industry

DOE acceptance of commercial mixed waste -- Studies are under way

Energy Technology Data Exchange (ETDEWEB)

Plummer, T.L. [Dept. of Energy, Washington, DC (United States). Technical Support Program; Owens, C.M. [Idaho National Engineering Lab., Idaho Falls, ID (United States). National Low-Level Waste Management Program

1993-03-01

The topic of the Department of Energy acceptance of commercial mixed waste at DOE facilities has been proposed by host States and compact regions that are developing low-level radioactive waste disposal facilities. States support the idea of DOE accepting commercial mixed waste because (a) very little commercial mixed waste is generated compared to generation by DOE facilities (Department of Energy--26,300 cubic meters annually vs. commercial--3400 cubic meters annually); (b) estimated costs for commercial disposal are estimated to be $15,000 to $40,000 per cubic foot; (c) once treatment capability becomes available, 70% of the current levels of commercial mixed waste will be eliminated, (d) some State laws prohibit the development of mixed waste disposal facilities in their States; (e) DOE is developing a nationwide strategy that will include treatment and disposal capacity for its own mixed waste and the incremental burden on the DOE facilities would be minuscule, and (6) no States are developing mixed waste disposal facilities. DOE senior management has repeatedly expressed willingness to consider investigating the feasibility of DOE accepting commercial mixed waste. In January 1991, Leo Duffy of the Department of energy met with members of the Low-Level Radioactive Waste Forum, which led to an agreement to explore such an arrangement. He stated that this seems like a cost-effective way to solve commercial mixed waste management problems.

Public acceptance in radioactive waste management

International Nuclear Information System (INIS)

Diaconu, Stela; Covreag, Ilinca

2008-01-01

Radioactive waste, unavoidable by-products of economically developed societies, arises from the production of energy by nuclear fission reactors as well as from medical, research and industrial applications of radioactive materials. The main objective of radioactive waste management is the safety as well the protection of public health and the environment. The first approach for the disposal of radioactive waste was based on the traditional 'decide, announce and defend' model, focused almost exclusively on technical content. In spite of the significant technical progress that would ensure long-term safety, the rate of progress towards implementing such solutions has been slower than expected, partly attributable to an earlier technical optimism and to an underestimation of the societal and political dimensions. It is now broadly recognized that radioactive waste management involves both technical and societal dimensions which cannot be dissociated. Because of changes in society's decision-making environment and heightened public sensitivity to all matters connected with environmental protection, nuclear power, radioactivity, and especially radioactive waste, any decision regarding whether, when and how to implement waste management solutions will typically require thorough public examination and the involvement of many relevant stakeholders. The building of a long-term relationship with the local communities and the waste management facility is one of the most important contributors to sustainable radioactive waste management solutions. A new approach in now in place at international level, based on 'engage, interact and co-operate', for which both technical and societal issues are to be reconciled. That means that the involvement of all interested parties in the decision-making process is a condition for a successful and publicly acceptable implementation of such a project. A central role in the public acceptance of nuclear technologies play the management and

PHYSICAL PROPERTY MEASUREMENTS OF LABORATORY PREPARED SALTSTONE GROUT

Energy Technology Data Exchange (ETDEWEB)

Hansen, E.; Cozzi, A.; Edwards, T.

2014-05-05

The Saltstone Production Facility (SPF) built two new Saltstone Disposal Units (SDU), SDU 3 and SDU 5, in 2013. The variable frequency drive (VFD) for the grout transfer hose pump tripped due to high current demand by the motor during the initial radioactive saltstone transfer to SDU 5B on 12/5/2013. This was not observed during clean cap processing on July 5, 2013 to SDU 3A, which is a slightly longer distance from the SPF than is SDU 5B. Saltstone Design Authority (SDA) is evaluating the grout pump performance and capabilities to transfer the grout processed in SPF to SDU 3/5. To assist in this evaluation, grout physical properties are required. At this time, there are no rheological data from the actual SPF so the properties of laboratory prepared samples using simulated salt solution or Tank 50 salt solution will be measured. The physical properties of grout prepared in the laboratory with de-ionized water (DI) and salt solutions were obtained at 0.60 and 0.59 water to premix (W/P) ratios, respectively. The yield stress of the DI grout was greater than any salt grout. The plastic viscosity of the DI grout was lower than all of the salt grouts (including salt grout with admixture). When these physical data were used to determine the pressure drop and fluid horsepower for steady state conditions, the salt grouts without admixture addition required a higher pressure drop and higher fluid horsepower to transport. When 0.00076 g Daratard 17/g premix was added, both the pressure drop and fluid horsepower were below that of the DI grout. Higher concentrations of Daratard 17 further reduced the pressure drop and fluid horsepower. The uncertainty in the single point Bingham Plastic parameters is + 4% of the reported values and is the bounding uncertainty. Two different mechanical agitator mixing protocols were followed for the simulant salt grout, one having a total mixing time of three minutes and the other having a time of 10 minutes. The Bingham Plastic parameters

HOW TO DEAL WITH WASTE ACCEPTANCE UNCERTAINTY USING THE WASTE ACCEPTANCE CRITERIA FORECASTING AND ANALYSIS CAPABILITY SYSTEM (WACFACS)

Energy Technology Data Exchange (ETDEWEB)

Redus, K. S.; Hampshire, G. J.; Patterson, J. E.; Perkins, A. B.

2002-02-25

The Waste Acceptance Criteria Forecasting and Analysis Capability System (WACFACS) is used to plan for, evaluate, and control the supply of approximately 1.8 million yd3 of low-level radioactive, TSCA, and RCRA hazardous wastes from over 60 environmental restoration projects between FY02 through FY10 to the Oak Ridge Environmental Management Waste Management Facility (EMWMF). WACFACS is a validated decision support tool that propagates uncertainties inherent in site-related contaminant characterization data, disposition volumes during EMWMF operations, and project schedules to quantitatively determine the confidence that risk-based performance standards are met. Trade-offs in schedule, volumes of waste lots, and allowable concentrations of contaminants are performed to optimize project waste disposition, regulatory compliance, and disposal cell management.

HOW TO DEAL WITH WASTE ACCEPTANCE UNCERTAINTY USING THE WASTE ACCEPTANCE CRITERIA FORECASTING AND ANALYSIS CAPABILITY SYSTEM (WACFACS)

International Nuclear Information System (INIS)

Redus, K. S.; Hampshire, G. J.; Patterson, J. E.; Perkins, A. B.

2002-01-01

The Waste Acceptance Criteria Forecasting and Analysis Capability System (WACFACS) is used to plan for, evaluate, and control the supply of approximately 1.8 million yd3 of low-level radioactive, TSCA, and RCRA hazardous wastes from over 60 environmental restoration projects between FY02 through FY10 to the Oak Ridge Environmental Management Waste Management Facility (EMWMF). WACFACS is a validated decision support tool that propagates uncertainties inherent in site-related contaminant characterization data, disposition volumes during EMWMF operations, and project schedules to quantitatively determine the confidence that risk-based performance standards are met. Trade-offs in schedule, volumes of waste lots, and allowable concentrations of contaminants are performed to optimize project waste disposition, regulatory compliance, and disposal cell management

Waste acceptance and waste loading for vitrified Oak Ridge tank waste

International Nuclear Information System (INIS)

Harbour, J.R.; Andrews, M.K.

1997-01-01

The Office of Science and Technology of the DOE has funded a joint project between the Oak Ridge National Laboratory (ORNL) and the Savannah River Technology Center (SRTC) to evaluate vitrification and grouting for the immobilization of sludge from ORNL tank farms. The radioactive waste is from the Gunite and Associated Tanks (GAAT), the Melton Valley Storage Tanks (MVST), the Bethel Valley Evaporator Service Tanks (BVEST), and the Old Hydrofractgure Tanks (OHF). Glass formulation development for sludge from these tanks is discussed in an accompanying article for this conference (Andrews and Workman). The sludges contain transuranic radionuclides at levels which will make the glass waste form (at reasonable waste loadings) TRU. Therefore, one of the objectives for this project was to ensure that the vitrified waste form could be disposed of at the Waste Isolation Pilot Plant (WIPP). In order to accomplish this, the waste form must meet the WIPP Waste Acceptance Criteria (WAC). An alternate pathway is to send the glass waste forms for disposal at the Nevada Test Site (NTS). A sludge waste loading in the feed of 6 wt percent will lead to a waste form which is non-TRU and could potentially be disposed of at NTS. The waste forms would then have to meet the requirements of the NTS WAC. This paper presents SRTC''s efforts at demonstrating that the glass waste form produced as a result of vitrification of ORNL sludge will meet all the criteria of the WIPP WAC or NTS WAC

Managing Hanford Site solid waste through strict acceptance criteria

International Nuclear Information System (INIS)

Jasen, W.G.; Pierce, R.D.; Willis, N.P.

1993-02-01

Various types of waste have been generated during the 50-year history of the Hanford Site. Regulatory changes in the last 20 years have provided the emphasis for better management of these wastes. Interpretations of the Atomic Energy Act of 1954 (AEA) and the Resource Conservation and Recovery Act of 1976 (RCRA) have led to the definition of a group of wastes called radioactive mixed wastes (RMW). As a result of the radioactive and hazardous properties of these wastes, strict management programs have been implemented for the management of these wastes. Solid waste management is accomplished through a systems performance approach to waste management that used best-demonstrated available technology (BDAT) and best management practices. The solid waste program at the Hanford Site strives to integrate all aspects of management relative to the treatment, storage and disposal (TSD) of solid waste. Often there are many competing and important needs. It is a difficult task to balance these needs in a manner that is both equitable and productive. Management science is used to help the process of making decisions. Tools used to support the decision making process include five-year planning, cost estimating, resource allocation, performance assessment, waste volume forecasts, input/output models, and waste acceptance criteria. The purpose of this document is to describe how one of these tools, waste acceptance criteria, has helped the Hanford Site manage solid wastes

Environmental Restoration Disposal Facility Waste Acceptance Criteria

International Nuclear Information System (INIS)

Dronen, V.R.

1998-06-01

The Hanford Site is operated by the U. S. Department of Energy (DOE) with a primary mission of environmental cleanup and restoration. The Environmental Restoration Disposal Facility (ERDF) is an integral part of the DOE environmental restoration effort at the Hanford Site. The purpose of this document is to establish the ERDF waste acceptance criteria for disposal of materials resulting from Hanford Site cleanup activities. Definition of and compliance with the requirements of this document will enable implementation of appropriate measures to protect human health and the environment, ensure the integrity of the ERDF liner system, facilitate efficient use of the available space in the ERDF, and comply with applicable environmental regulations and DOE orders. To serve this purpose, the document defines responsibilities, identifies the waste acceptance process, and provides the primary acceptance criteria and regulatory citations to guide ERDF users. The information contained in this document is not intended to repeat or summarize the contents of all applicable regulations

High-Level Waste System Process Interface Description

International Nuclear Information System (INIS)

D'Entremont, P.D.

1999-01-01

The High-Level Waste System is a set of six different processes interconnected by pipelines. These processes function as one large treatment plant that receives, stores, and treats high-level wastes from various generators at SRS and converts them into forms suitable for final disposal. The three major forms are borosilicate glass, which will be eventually disposed of in a Federal Repository, Saltstone to be buried on site, and treated water effluent that is released to the environment

Special Analysis: Revised 14C Disposal Limits for the Saltstone Disposal Facility

International Nuclear Information System (INIS)

Kaplan, D.I.

2004-01-01

The Saltstone Special Analysis calculated a limit for 14C based on the atmospheric pathway of 52 pCi/mL using some very conservative assumptions. This was compared to the estimated Low Curie Salt concentration of 0.45 pCi/mL and since the limit was two orders of magnitude greater than the estimated concentration, the decision was made that no further analysis was needed. The 14C concentration in Tank 41 has been found to be much greater than the estimated concentration and to exceed the limit derived in the Special Analysis. A rigorous analysis of the release of 14C via the air pathway that considers the chemical effects of the Saltstone system has shown that the flux of 14C is significantly less than that assumed in the Special Analysis. The net result is an inventory limit for 14C that is significantly higher than that derived in the Special Analysis that will also meet the performance objectives of DOE Order 435.1

Study of waste acceptance criteria for low-level radioactive waste from medical, industrial, and research facilities (Contract research)

International Nuclear Information System (INIS)

Koibuchi, Hiroto; Dohi, Terumi; Ishiguro, Hideharu; Hayashi, Masaru; Senda, Masaki

2008-12-01

Japan Atomic Energy Agency (JAEA) is supposed to draw up the plan for the disposal program of the very low-level radioactive waste and low-level radioactive waste generated from medical, industrial and research facilities. For instance, there are these facilities in JAEA, universities, private companies, and so on. JAEA has to get to know about the waste and its acceptance of other institutions described above because it is important for us to hold the licenses for the disposal program regarding safety assessment. This report presents the basic data concerning radioactive waste of research institutes etc. except RI waste, domestic and foreign information related to acceptance criteria for disposal of the low-level radioactive waste, the current status of foreign medical waste management, waste acceptance, and such. In this report, Japan's acceptance criteria were summarized on the basis of present regulation. And, the criteria of foreign countries, United States, France, United Kingdom and Spain, were investigated by survey of each reference. In addition, it was reported that the amount of waste from laboratories etc. for near-surface disposal and their characterization in our country. The Subjects of future work: the treatment of hazardous waste, the problem of the double-regulation (the Nuclear Reactor Regulation Law and the Law Concerning Prevention from Radiation Hazards due to Radioisotopes and Others) and the possession of waste were discussed here. (author)

PORFLOW Simulations Supporting Saltstone Disposal Unit Design Optimization

Energy Technology Data Exchange (ETDEWEB)

Flach, G. P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hang, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Taylor, G. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-12-10

SRNL was requested by SRR to perform PORFLOW simulations to support potential cost-saving design modifications to future Saltstone Disposal Units in Z-Area (SRR-CWDA-2015-00120). The design sensitivity cases are defined in a modeling input specification document SRR-CWDA-2015-00133 Rev. 1. A high-level description of PORFLOW modeling and interpretation of results are provided in SRR-CWDA-2015-00169. The present report focuses on underlying technical issues and details of PORFLOW modeling not addressed by the input specification and results interpretation documents. Design checking of PORFLOW modeling is documented in SRNL-L3200-2015-00146.

Analysis of SRP waste streams for waste tank certification

International Nuclear Information System (INIS)

Coleman, C.J.

1989-01-01

The Savannah River Plant (SRP) will apply for certification from the State of South Carolina to operate the SRP High-Level Waste Tanks. The permit application will be submitted as a RCRA Part B, Volume 16, entitled ''RCRA Part B Application For the F and H-Area Radioactive Waste Farm.'' RCRA regulations require that influent and effluent streams of hazardous waste sites be characterized to obtain an operating permit. The Waste Management Technology Department requested ADD to determine 21 components (including pH and weight percent solids) in the current influent streams to SRP High-Level Waste Tanks. The analyses will be used to supplement existing data on the composition of High-Level Waste. Effluent streams, which will feed Saltstone and the DWPF, will be analyzed when they are produced. This report contains the data obtained from analyzing key influent streams to SRP High-Level Waste Tanks. The precision of the data and the analytical methods that were used are also discussed

Waste acceptance criteria for the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

1996-04-01

The Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC), DOE/WIPP-069, was initially developed by a U.S. Department of Energy (DOE) Steering Committee to provide performance requirements to ensure public health and safety as well as the safe handling of transuranic (TRU) waste at the WIPP. This revision updates the criteria and requirements of previous revisions and deletes those which were applicable only to the test phase. The criteria and requirements in this document must be met by participating DOE TRU Waste Generator/Storage Sites (Sites) prior to shipping contact-handled (CH) and remote-handled (RH) TRU waste forms to the WIPP. The WIPP Project will comply with applicable federal and state regulations and requirements, including those in Titles 10, 40, and 49 of the Code of Federal Regulations (CFR). The WAC, DOE/WIPP-069, serves as the primary directive for assuring the safe handling, transportation, and disposal of TRU wastes in the WIPP and for the certification of these wastes. The WAC identifies strict requirements that must be met by participating Sites before these TRU wastes may be shipped for disposal in the WIPP facility. These criteria and requirements will be reviewed and revised as appropriate, based on new technical or regulatory requirements. The WAC is a controlled document. Revised/changed pages will be supplied to all holders of controlled copies

DISPOSABLE CANISTER WASTE ACCEPTANCE CRITERIA

Energy Technology Data Exchange (ETDEWEB)

R.J. Garrett

2001-07-30

The purpose of this calculation is to provide the bases for defining the preclosure limits on radioactive material releases from radioactive waste forms to be received in disposable canisters at the Monitored Geologic Repository (MGR) at Yucca Mountain. Specifically, this calculation will provide the basis for criteria to be included in a forthcoming revision of the Waste Acceptance System Requirements Document (WASRD) that limits releases in terms of non-isotope-specific canister release dose-equivalent source terms. These criteria will be developed for the Department of Energy spent nuclear fuel (DSNF) standard canister, the Multicanister Overpack (MCO), the naval spent fuel canister, the High-Level Waste (HLW) canister, the plutonium can-in-canister, and the large Multipurpose Canister (MPC). The shippers of such canisters will be required to demonstrate that they meet these criteria before the canisters are accepted at the MGR. The Quality Assurance program is applicable to this calculation. The work reported in this document is part of the analysis of DSNF and is performed using procedure AP-3.124, Calculations. The work done for this analysis was evaluated according to procedure QAP-2-0, Control of Activities, which has been superseded by AP-2.21Q, Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities. This evaluation determined that such activities are subject to the requirements of DOE/RW/0333P, Quality Assurance Requirements and Description (DOE 2000). This work is also prepared in accordance with the development plan titled Design Basis Event Analyses on DOE SNF and Plutonium Can-In-Canister Waste Forms (CRWMS M&O 1999a) and Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages (CRWMS M&O 2000d). This calculation contains no electronic data applicable to any electronic data management system.

SENSITIVITY ANALYSIS FOR SALTSTONE DISPOSAL UNIT COLUMN DEGRADATION ANALYSES

Energy Technology Data Exchange (ETDEWEB)

Flach, G.

2014-10-28

PORFLOW related analyses supporting a Sensitivity Analysis for Saltstone Disposal Unit (SDU) column degradation were performed. Previous analyses, Flach and Taylor 2014, used a model in which the SDU columns degraded in a piecewise manner from the top and bottom simultaneously. The current analyses employs a model in which all pieces of the column degrade at the same time. Information was extracted from the analyses which may be useful in determining the distribution of Tc-99 in the various SDUs throughout time and in determining flow balances for the SDUs.

Physical system requirements - Accept waste

International Nuclear Information System (INIS)

1992-08-01

The Nuclear Waste Policy Act (NWPA) assigned to the Department of Energy (DOE) the responsibility for managing the disposal of spent nuclear fuel and high-level radioactive waste and established the Office of Civilian Radioactive Waste Management (OCRWM) for that purpose. The Secretary of Energy, in his November 1989 report to Congress (DOE/RW-0247), announced new initiatives for the conduct of the Civilian Radioactive Waste Management (CRWM) program. One of these initiatives was to establish improved management structure and procedures. In response, OCRWM performed a management study and the OCRWM Director subsequently issued the Management Systems improvement Strategy (MSIS) on August 10, 1990, calling for a rigorous implementation of systems engineering principles with a special emphasis on functional analysis. The functional analysis approach establishes a framework for integrating the program management efforts with the technical requirements analysis into a single, unified, and consistent program. This approach recognizes that just as the facilities and equipment comprising the physical waste management system must perform certain functions, so must certain programmatic and management functions be performed within the program in order to successfully bring the physical system into being. Thus, a comprehensive functional analysis effort has been undertaken which is intended to: Identify the functions that must be performed to fulfill the waste disposal mission; Identify the corresponding requirements imposed on each of the functions; and Identify the conceptual architecture that will be used to satisfy the requirements. The principal purpose of this requirements document is to present the results that were obtained from the conduct of a functional analysis effort for the Accept Waste mission

Documentation of acceptable knowledge for Los Alamos National Laboratory Plutonium Facility TRU waste stream

International Nuclear Information System (INIS)

Montoya, A.J.; Gruetzmacher, K.M.; Foxx, C.L.; Rogers, P.Z.

1998-03-01

Characterization of transuranic waste from the LANL Plutonium Facility for certification and transportation to WIPP includes the use of acceptable knowledge as specified in the WIPP Quality Assurance Program Plan. In accordance with a site specific procedure, documentation of acceptable knowledge for retrievably stored and currently generated transuranic waste streams is in progress at LANL. A summary overview of the TRU waste inventory is complete and documented in the Sampling Plan. This document also includes projected waste generation, facility missions, waste generation processes, flow diagrams, times, and material inputs. The second part of acceptable knowledge documentation consists of assembling more detailed acceptable knowledge information into auditable records and is expected to require several years to complete. These records for each waste stream must support final assignment of waste matrix parameters, EPA hazardous waste numbers, and radionuclide characterization. They must also include a determination whether waste streams are defense waste streams for compliance with the WIPP Land Withdrawal Act. The LANL Plutonium Facility's mission is primarily plutonium processing in basic special nuclear material (SNM) research activities to support national defense and energy programs. It currently has about 100 processes ranging from SNM recovery from residues to development of plutonium 238 heat sources for space applications. Its challenge is to characterize and certify waste streams from such diverse and dynamic operations using acceptable knowledge. This paper reports the progress on the certification of the first of these waste streams to the WIPP WAC

Characteristics of radioactive waste forms conditioned for storage and disposal: Guidance for the development of waste acceptance criteria

International Nuclear Information System (INIS)

1983-04-01

This report attempts to review the characteristics of the individual components of the waste package, i.e. the waste form and the container, in order to formulate, where appropriate, quidelines for the development of practical waste acceptance criteria. Primarily the criteria for disposal are considered, but if more stringent criteria are expected to be necessary for storage or transportation prior to the disposal, these will be discussed. The report will also suggest test areas which will aid the development of the final waste acceptance criteria

Acceptance criteria for radioactive waste deposition

International Nuclear Information System (INIS)

Rzyski, B.M.

1989-01-01

The disposal of low-and intermediate level radioactive waste in either shallow ground or rock cavities must be subjected to special guidelines which are used by national authorities and implementing bodies when establishing and regulating respositories. These informations are given by the acceptance criteria and will depend on specific site conditions and optmized procedures. (author) [pt

Long-Term Safety Analysis of Baldone Radioactive Waste Repository and Updating of Waste Acceptance Criteria

International Nuclear Information System (INIS)

2001-12-01

The main objective of the project was to provide advice to the Latvian authorities on the safety enhancements and waste acceptance criteria for near surface radioactive waste disposal facilities of the Baldone repository. The project included the following main activities: Analysis of the current status of the management of radioactive waste in Latvia in general and, at the Baldone repository in particular Development of the short and long-term safety analysis of the Baldone repository, including: the planned increasing of capacity for disposal and long term storage, the radiological analysis for the post-closure period Development of the Environment Impact Statement, for the new foreseen installations, considering the non radiological components Proposal of recommendations for future updating of radioactive waste acceptance criteria Proposal of recommendations for safety upgrades to the facility. The work programme has been developed in phases and main tasks as follows. Phase 0: Project inception, Phase 1: Establishment of current status, plans and practices (Legislation, regulation and standards, Radioactive waste management, Waste acceptance criteria), Phase 2: Development of future strategies for long-term safety management and recommendations for safety enhancements. The project team found the general approach use at the installation, the basic design and the operating practices appropriate to international standards. Nevertheless, a number of items subject to potential improvements were also identified. These upgrading recommendations deal with general aspects of the management (mainly storage versus disposal of long-lived sources), site and environmental surveillance, packaging (qualification of containers, waste characterization requirements), the design of an engineered cap and strategies for capping. (author)

Waste Acceptance Decisions and Uncertainty Analysis at the Oak Ridge Environmental Management Waste Management Facility

International Nuclear Information System (INIS)

Redus, K. S.; Patterson, J. E.; Hampshire, G. L.; Perkins, A. B.

2003-01-01

The Waste Acceptance Criteria (WAC) Attainment Team (AT) routinely provides the U.S. Department of Energy (DOE) Oak Ridge Operations with Go/No-Go decisions associated with the disposition of over 1.8 million yd3 of low-level radioactive, TSCA, and RCRA hazardous waste. This supply of waste comes from 60+ environmental restoration projects over the next 15 years planned to be dispositioned at the Oak Ridge Environmental Management Waste Management Facility (EMWMF). The EMWMF WAC AT decision making process is accomplished in four ways: (1) ensure a clearly defined mission and timeframe for accomplishment is established, (2) provide an effective organization structure with trained personnel, (3) have in place a set of waste acceptance decisions and Data Quality Objectives (DQO) for which quantitative measures are required, and (4) use validated risk-based forecasting, decision support, and modeling/simulation tools. We provide a summary of WAC AT structure and performance. We offer suggestions based on lessons learned for effective transfer to other DOE

Waste Acceptance Decisions and Uncertainty Analysis at the Oak Ridge Environmental Management Waste Management Facility

Energy Technology Data Exchange (ETDEWEB)

Redus, K. S.; Patterson, J. E.; Hampshire, G. L.; Perkins, A. B.

2003-02-25

The Waste Acceptance Criteria (WAC) Attainment Team (AT) routinely provides the U.S. Department of Energy (DOE) Oak Ridge Operations with Go/No-Go decisions associated with the disposition of over 1.8 million yd3 of low-level radioactive, TSCA, and RCRA hazardous waste. This supply of waste comes from 60+ environmental restoration projects over the next 15 years planned to be dispositioned at the Oak Ridge Environmental Management Waste Management Facility (EMWMF). The EMWMF WAC AT decision making process is accomplished in four ways: (1) ensure a clearly defined mission and timeframe for accomplishment is established, (2) provide an effective organization structure with trained personnel, (3) have in place a set of waste acceptance decisions and Data Quality Objectives (DQO) for which quantitative measures are required, and (4) use validated risk-based forecasting, decision support, and modeling/simulation tools. We provide a summary of WAC AT structure and performance. We offer suggestions based on lessons learned for effective transfer to other DOE.

Scientific, institutional, regulatory, political, and public acceptance of the waste isolation pilot plant transuranic waste repository

International Nuclear Information System (INIS)

Eriksson, L.G.

2000-01-01

The recent successful certification and opening of a first-of-a-kind, deep geological repository for safe disposal of long-lived, transuranic radioactive waste (TRUW) at the Waste Isolation Pilot Plant (WIPP) site, New Mexico, United States of America (USA), embody both long-standing local and wide-spread, gradually achieved, scientific, institutional, regulatory, political, and public acceptance. The related historical background and development are outlined and the main contributors to the successful siting, certification, and acceptance of the WIPP TRUW repository, which may also serve as a model to success for other radioactive waste disposal programs, are described. (author)

Institutional innovation to generate the public acceptance of radioactive waste disposal

International Nuclear Information System (INIS)

Kemp, R.

1991-01-01

Contrasting experiences of public acceptance of radioactive waste disposal are compared for the United Kingdom, France, Sweden and Canada. The disparity between scientifically assessed and publicly perceived levels of risk is noted. The author argues that the form of decision-making process is more important to public acceptance of radioactive waste disposal than the technology of disposal. Public risk perception can be altered by procedures employed in planning, negotiation and consultation. Precisely what constitutes acceptable risk does vary from country to country, and differences in institutional responses and innovation are particularly highlighted. (UK)

Derivation of Waste Acceptance Criteria for Low and Intermediate Level Waste in Surface Disposal Facility

International Nuclear Information System (INIS)

Gagner, L.; Voinis, S.

2000-01-01

In France, low- and intermediate-level radioactive wastes are disposed in a near-surface facility, at Centre de l'Aube disposal facility. This facility, which was commissioned in 1992, has a disposal capacity of one million cubic meters, and will be operated up to about 2050. It took over the job from Centre de la Manche, which was commissioned in 1969 and shut down in 1994, after having received about 520,000 cubic meters of wastes. The Centre de l'Aube disposal facility is designed to receive a many types of waste produced by nuclear power plants, reprocessing, decommissioning, as well as by the industry, hospitals and armed forces. The limitation of radioactive transfer to man and the limitation of personnel exposure in all situations considered plausible require limiting the total activity of the waste disposed in the facility as well as the activity of each package. The paper presents how ANDRA has derived the activity-related acceptance criteria, based on the safety analysis. In the French methodology, activity is considered as end-point for deriving the concentration limits per package, whereas it is the starting point for deriving the total activity limits. For the concentration limits (called here LMA) the approach consists of five steps: the determination of radionuclides important for safety with regards to operational and long-term safety, the use of relevant safety scenarios as a tool to derive quantitative limits, the setting of dose constraint per situation associated with scenarios, the setting of contribution factor per radionuclide, and the calculation of concentration activity limits. An exhaustive survey has been performed and has shown that the totality of waste packages which should be delivered by waste generators are acceptable in terms of activity limits in the Centre de l'Aube. Examples of concentration activity limits derived from this methodology are presented. Furthermore those limits have been accepted by the French regulatory body and

Acceptance of non-fuel assembly hardware by the Federal Waste Management System

International Nuclear Information System (INIS)

1990-03-01

This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high-level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. 14 refs., 12 figs., 43 tabs

WRAP low level waste restricted waste management (LLW RWM) glovebox acceptance test report

International Nuclear Information System (INIS)

Leist, K.J.

1997-01-01

On April 22, 1997, the Low Level Waste Restricted Waste Management (LLW RWM) glovebox was tested using acceptance test procedure 13027A-87. Mr. Robert L. Warmenhoven served as test director, Mr. Kendrick Leist acted as test operator and test witness, and Michael Lane provided miscellaneous software support. The primary focus of the glovebox acceptance test was to examine glovebox control system interlocks, operator Interface Unit (OIU) menus, alarms, and messages. Basic drum port and lift table control sequences were demonstrated. OIU menus, messages, and alarm sequences were examined, with few exceptions noted. Barcode testing was bypassed, due to the lack of installed equipment as well as the switch from basic reliance on fixed bar code readers to the enhanced use of portable bar code readers. Bar code testing was completed during performance of the LLW RWM OTP. Mechanical and control deficiencies were documented as Test Exceptions during performance of this Acceptance Test. These items are attached as Appendix A to this report

Management of radioactive waste in Belgium: ONDRAF/NIRAS and Belgoprocess as major actors of the waste acceptance system

International Nuclear Information System (INIS)

Zaelen, Gunter van; Verheyen, Annick

2007-01-01

The management of radioactive waste in Belgium is undertaken by the national agency for radioactive waste and enriched fissile materials, ONDRAF/NIRAS, and its industrial partner Belgoprocess. ONDRAF/NIRAS has set up a management system designed to guarantee that the general public and the environment are protected against the potential hazards arising from radioactive waste. Belgoprocess is a private company, founded in 1984 and located in Dessel, Belgium. It is a subsidiary of ONDRAF/NIRAS and its activities focus on the safe processing and storage of radioactive waste. The management system of ONDRAF/NIRAS includes two aspects: a) an integrated system and b) an acceptance system. The integrated system covers all aspects of management ranging from the origin of waste to its transport, processing, interim storage and long-term management. The safety of radioactive waste management not only depends on the quality of the design and construction of the processing, temporary storage or disposal infrastructure, but also on the quality of the waste accepted by ONDRAF/NIRAS. In order to be manage d safely, both in the short and the long term, the waste transferred to ONDRAF/NIRAS must meet certain specific requirements. To that end, ONDRAF/NIRAS has developed an acceptance system. (authors)

Public acceptance for centralized storage and repositories of low-level waste session (Panel)

Energy Technology Data Exchange (ETDEWEB)

Lutz, H.R.

1995-12-31

Participants from various parts of the world will provide a summary of their particular country`s approach to low-level waste management and the cost of public acceptance for low-level waste management facilities. Participants will discuss the number, geographic location, and type of low-level waste repositories and centralized storage facilities located in their countries. Each will discuss the amount, distribution, and duration of funds to gain public acceptance of these facilities. Participants will provide an estimated $/meter for centralized storage facilities and repositories. The panel will include a brief discussion about the ethical aspects of public acceptance costs, approaches for negotiating acceptance, and lessons learned in each country. The audience is invited to participate in the discussion.

Preliminary waste acceptance criteria for the ICPP spent fuel and waste management technology development program

International Nuclear Information System (INIS)

Taylor, L.L.; Shikashio, R.

1993-09-01

The purpose of this document is to identify requirements to be met by the Producer/Shipper of Spent Nuclear Fuel/High-LeveL Waste SNF/HLW in order for DOE to be able to accept the packaged materials. This includes defining both standard and nonstandard waste forms

Summary of research and development activities in support of waste acceptance criteria for WIPP

International Nuclear Information System (INIS)

Hunter, T.O.

1979-11-01

The development of waste acceptance criteria for the Waste Isolation Pilot Plant (WIPP) is summarized. Specifications for acceptable waste forms are included. Nine program areas are discussed. They are: TRU characterization, HLW interactions, thermal/structural interactions, nuclide migration, permeability, brine migration, borehole plugging, operation/design support, and instrumentation development. Recommendations are included

Supplemental Immobilization of Hanford Low-Activity Waste: Cast Stone Screening Tests

Energy Technology Data Exchange (ETDEWEB)

Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Piepel, Gregory F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lindberg, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Heasler, Patrick G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mercier, Theresa M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Russell, Renee L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cozzi, Alex [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Daniel, William E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Eibling, Russell E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hansen, E. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Reigel, Marissa M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Swanberg, David J. [Washington River Protection Solutions (WRPS), Aiken, SC (United States)

2013-09-30

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 southeastern Washington State. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the wastes and immobilize them in a glass waste form. The WTP includes a pretreatment facility to separate the wastes into a small volume of high-level waste (HLW) containing most of the radioactivity and a larger volume of low-activity waste (LAW) containing most of the nonradioactive chemicals. The HLW will be converted to glass in the HLW vitrification facility for ultimate disposal at an offsite federal repository. At least a portion (~35%) of the LAW will be converted to glass in the LAW vitrification facility and will be disposed of onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize the wastes destined for each facility. However, a second LAW immobilization facility will be needed for the expected volume of LAW requiring immobilization. A cementitious waste form known as Cast Stone 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. Further, the waste form must be tested to provide the technical basis for understanding the long-term performance of the waste form in the disposal environment. These waste form performance data are needed to support risk assessment and performance assessment (PA) analyses of the long-term environmental impact of the waste disposal in the IDF

Acceptance of failed SNF [spent nuclear fuel] assemblies by the Federal Waste Management System

International Nuclear Information System (INIS)

1990-03-01

This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. This document discusses acceptance of failed spent fuel assemblies by the Federal Waste Management System. 18 refs., 7 figs., 25 tabs

Properties of slag concrete for low-level waste containment

International Nuclear Information System (INIS)

Langton, C.A.; Wong, P.B.

1991-01-01

Ground granulated blast furnace slag was incorporated in the concrete mix used for construction of low-level radioactive waste disposal vaults. The vaults were constructed as six 100 x 100 x 25 ft cells with each cell sharing internal walls with the two adjacent cells. The vaults were designed to contain a low-level radioactive wasteform called saltstone and to isolate the saltstone from the environment until the landfill is closed. Closure involves backfilling with native soil, installation of clay cap, and run-off control. The design criteria for the slag-substituted concrete included compressive strength, 4000 psi after 28 days; slump, 6 inch; permeability, less than 10 -7 cm/sec; and effective nitrate, chromium and technetium diffusivities of 10 -8 , 10 -12 and 10 -12 cm 2 /sec, respectively. The reducing capacity of the slag resulted in chemically reducing Cr +6 to Cr +3 and Tc +7 to Tc +4 and subsequent precipitation of the respective hydroxides in the alkaline pore solution. Consequently, the concrete vault enhances containment of otherwise mobile waste ions and contributes to the overall protection of the groundwater at the disposal site

From waste packages acceptance criteria to waste packages acceptance process at the Centre de l'Aube disposal facility

International Nuclear Information System (INIS)

Dutzer, M.

2003-01-01

The Centre de l'Aube disposal facility has now been operated for 10 years. At the end of 2001, about 124,000 m3 of low and intermediate level short lived waste packages, representing 180,000 packages, have been disposed, for a total capacity of 1,000,000 m3. The flow of waste packages is now between 12 and 15,000 m3 per year, that is one third of the flow that was taken into account for the design of the repository. It confirms the efforts by waste generators to minimise waste production. This flow represents 25 to 30,000 packages, 50% are conditioned into the compaction facility of the repository, so that 17,000 packages are disposed per year. 54 disposal vaults have been closed. In 1996-1999, the safety assessment of the repository have been reviewed, taking into account the experience of operation. This assessment was investigated by the regulatory body and, subsequently, a so-called 'definitive license' to operate was granted to ANDRA on September 2, 1999 with updated licensing requirements. Another review will be performed in 2004. To ensure a better consistency with the safety assessment of the facility, Andra issued new technical requirements for waste packages at the end of 2000. Discussions with waste generators also showed that the waste package acceptance process should be improved to provide a more precise definition of operational criteria to comply with in waste conditioning facilities. Consequently, a new approach has been implemented since 2000. (orig.)

Acceptance of waste for disposal in the potential United States repository at Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Stahl, D.; Svinicki, K.

1996-01-01

This paper addresses the process for the acceptance of waste into the waste management system (WMS) with a focus on the detailed requirements identified from the Waste Acceptance System Requirements Document. Also described is the recent dialogue between OCRWM and the Office of Environmental Management to resolve issues, including the appropriate interpretation and application of regulatory and system requirements to DOE-owned spent fuel. Some information is provided on the design of the repository system to aid the reader in understanding how waste that is accepted into the WMS is received and emplaced in the repository

Waste acceptance and impact ON D and D in Switzerland

International Nuclear Information System (INIS)

Maxeiner, Harald

2002-01-01

Harald Maxeiner described clearance and waste conditioning requirements in Switzerland, and their impacts on decommissioning: Although decommissioning of the first (oldest) reactor will not take place until 2009 at the earliest (hypothetical operating lifetime of 40 years), detailed decommissioning studies have to be carried out today, in order to demonstrate the feasibility of the technologies to be used and to determine anticipated costs (for the purpose of calculating financial contributions to a decommissioning fund). The studies are based on waste acceptance criteria and guidelines that apply to waste already in existence. The focus is on preparing inventories of activated and contaminated components and conditioning of these components. The basis for present and future conditioning of radioactive wastes, as well as for their interim storage and final disposal, is provided by the official guideline HSK R-14. According to this guideline, raw waste requires to be solidified (inter alia with cement) and the resulting waste product must: remain intact until final disposal, not be readily dispersible, be resistant to aqueous media, not be readily combustible, not contain any unnecessary voids, contain as little organic material as possible. The waste package containing the waste product must: constitute a further barrier to dispersion, outlast (at least) interim storage, be documented with details of manufacturing, composition, properties, be designed to resist corrosion using suitable materials, be characterised by a quality assurance program for raw waste, waste product and waste package. The only possible reasons for interim storage of waste without solidification are: decay storage followed by conventional waste management, if waste packages fulfil acceptance criteria for the final repository without further treatment, if, in the foreseeable future, an alternative conditioning method can be expected. The guidelines and acceptance criteria mentioned set strict

Acceptance criteria for disposal of radioactive waste in Romania

International Nuclear Information System (INIS)

Dogaru, D.

2001-01-01

In Romania the institutional radioactive waste are managed by National Institute of R and D for Physics and Nuclear Engineering. The institutional radioactive waste are collected, treated and conditioned at the Radioactive Waste Treatment Plant then transferred and disposed to the National Repository of Radioactive Waste at Baita Bihor. National Repository for Radioactive Waste is a long term storage facility. The repository is placed in a former worked out uranium ore mine, being excavated in the Bihor peak. The repository has been sited taking into account the known geological, hydrogeoloical, seismic and meteorological and mining properties of a uranium mining site. In the absence of an updated Safety Analysis Report, the maximum radioactive content permitted by the regulatory authority in the operation license is below the values reported for other engineered repositories in mine galleries. The paper presents the acceptance criteria for disposal of radioactive waste in National Repository for Radioactive Waste at Baita Bihor. (author)

Methods for verifying compliance with low-level radioactive waste acceptance criteria

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-09-01

This report summarizes the methods that are currently employed and those that can be used to verify compliance with low-level radioactive waste (LLW) disposal facility waste acceptance criteria (WAC). This report presents the applicable regulations representing the Federal, State, and site-specific criteria for accepting LLW. Typical LLW generators are summarized, along with descriptions of their waste streams and final waste forms. General procedures and methods used by the LLW generators to verify compliance with the disposal facility WAC are presented. The report was written to provide an understanding of how a regulator could verify compliance with a LLW disposal facility`s WAC. A comprehensive study of the methodology used to verify waste generator compliance with the disposal facility WAC is presented in this report. The study involved compiling the relevant regulations to define the WAC, reviewing regulatory agency inspection programs, and summarizing waste verification technology and equipment. The results of the study indicate that waste generators conduct verification programs that include packaging, classification, characterization, and stabilization elements. The current LLW disposal facilities perform waste verification steps on incoming shipments. A model inspection and verification program, which includes an emphasis on the generator`s waste application documentation of their waste verification program, is recommended. The disposal facility verification procedures primarily involve the use of portable radiological survey instrumentation. The actual verification of generator compliance to the LLW disposal facility WAC is performed through a combination of incoming shipment checks and generator site audits.

Methods for verifying compliance with low-level radioactive waste acceptance criteria

International Nuclear Information System (INIS)

1993-09-01

This report summarizes the methods that are currently employed and those that can be used to verify compliance with low-level radioactive waste (LLW) disposal facility waste acceptance criteria (WAC). This report presents the applicable regulations representing the Federal, State, and site-specific criteria for accepting LLW. Typical LLW generators are summarized, along with descriptions of their waste streams and final waste forms. General procedures and methods used by the LLW generators to verify compliance with the disposal facility WAC are presented. The report was written to provide an understanding of how a regulator could verify compliance with a LLW disposal facility's WAC. A comprehensive study of the methodology used to verify waste generator compliance with the disposal facility WAC is presented in this report. The study involved compiling the relevant regulations to define the WAC, reviewing regulatory agency inspection programs, and summarizing waste verification technology and equipment. The results of the study indicate that waste generators conduct verification programs that include packaging, classification, characterization, and stabilization elements. The current LLW disposal facilities perform waste verification steps on incoming shipments. A model inspection and verification program, which includes an emphasis on the generator's waste application documentation of their waste verification program, is recommended. The disposal facility verification procedures primarily involve the use of portable radiological survey instrumentation. The actual verification of generator compliance to the LLW disposal facility WAC is performed through a combination of incoming shipment checks and generator site audits

Risk analysis for new nuclear waste sites: Will it generate public acceptance?

International Nuclear Information System (INIS)

Inhaber, H.

1993-01-01

This report discusses public acceptance of radioactive waste facilities and what seems to be increasingly militant stances against such facilities. The role of risk assessment in possibly enhancing public acceptance is investigated

Liquid secondary waste: Waste form formulation and qualification

Energy Technology Data Exchange (ETDEWEB)

Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-07-31

The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, including Direct Feed Low Activity Waste (DFLAW) vitrification, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. The powdered salt waste form produced by the ETF will be replaced by a stabilized solidified waste form for disposal in Hanford’s Integrated Disposal Facility (IDF). Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the IDF. Waste form testing to support this plan is composed of work in the near term to provide data as input to a performance assessment (PA) for Hanford’s IDF. In 2015, three Hanford Liquid Secondary Waste simulants were developed based on existing and projected waste streams. Using these waste simulants, fourteen mixes of Hanford Liquid Secondary Waste were prepared and tested varying the waste simulant, the water-to-dry materials ratio, and the dry materials blend composition.1 In FY16, testing was performed using a simulant of the EMF process condensate blended with the caustic scrubber—from the Low Activity Waste (LAW) melter—, processed through the ETF. The initial EMF-16 simulant will be based on modeling efforts performed to determine the mass balance of the ETF for the DFLAW.2 The compressive strength of all of the mixes exceeded the target of 3.4 MPa (500 psi) to meet the requirements identified as potential IDF Waste Acceptance Criteria in Table 1 of the Secondary Liquid Waste Immobilization Technology Development Plan.3 The hydraulic properties of the waste forms tested (hydraulic conductivity

Quality assurance requirements and methods for high level waste package acceptability

International Nuclear Information System (INIS)

1992-12-01

This document should serve as guidance for assigning the necessary items to control the conditioning process in such a way that waste packages are produced in compliance with the waste acceptance requirements. It is also provided to promote the exchange of information on quality assurance requirements and on the application of quality assurance methods associated with the production of high level waste packages, to ensure that these waste packages comply with the requirements for transportation, interim storage and waste disposal in deep geological formations. The document is intended to assist both the operators of conditioning facilities and repositories as well as national authorities and regulatory bodies, involved in the licensing of the conditioning of high level radioactive wastes or in the development of deep underground disposal systems. The document recommends the quality assurance requirements and methods which are necessary to generate data for these parameters identified in IAEA-TECDOC-560 on qualitative acceptance criteria, and indicates where and when the control methods can be applied, e.g. in the operation or commissioning of a process or in the development of a waste package design. Emphasis is on the control of the process and little reliance is placed on non-destructive or destructive testing. Qualitative criteria, relevant to disposal of high level waste, are repository dependent and are not addressed here. 37 refs, 3 figs, 2 tabs

Reevaluating NIMBY: Evolving Public Fear and Acceptance in Siting a Nuclear Waste Facility

Energy Technology Data Exchange (ETDEWEB)

Jenkins-Smith, Hank C.; Silva, Carol L.; Nowlin, Matthew C.; deLozier, Grant (Dept. of Political Science, Univ. of Oklahoma, Norman, OK (United States))

2010-09-15

The not-in-my-backyard (NIMBY) syndrome has long been the focus of academic and policy research. We test several competing hypothesis concerning the sources of NIMBY sentiments, including demographics, proximity, political ideology and partisanship, and the unfolding policy process over time. To test these hypotheses we use survey data collected in New Mexico dealing with risk perceptions and acceptance related to the Waste Isolation Pilot Project (WIPP), a permanent storage site for radioactive waste located near Carlsbad, New Mexico. WIPP became operational and received its first shipment of waste on March 26, 1999. This study tracks the changes of risk perception and acceptance over a decade, using measures taken from 35 statewide surveys of New Mexico citizens spanning the 11-year period from fall 1990 to summer 2001. This time span includes periods before and after WIPP became operational. We find that acceptance of WIPP is greater among those in the most proximate counties to WIPP. Surprisingly, and contrary to expectations drawn from the broader literature, acceptance is also greater among those who live closest to the nuclear waste transportation route. We also find that ideology, partisanship, government approval and broader environmental concerns influence support for WIPP acceptance. Finally, the sequence of procedural steps taken toward formal approval of WIPP by government agencies proved to be important to public acceptance, the most significant being the opening of the WIPP facility itself

Reevaluating NIMBY: Evolving Public Fear and Acceptance in Siting a Nuclear Waste Facility

International Nuclear Information System (INIS)

Jenkins-Smith, Hank C.; Silva, Carol L.; Nowlin, Matthew C.; de Lozier, Grant

2010-09-01

The not-in-my-backyard (NIMBY) syndrome has long been the focus of academic and policy research. We test several competing hypothesis concerning the sources of NIMBY sentiments, including demographics, proximity, political ideology and partisanship, and the unfolding policy process over time. To test these hypotheses we use survey data collected in New Mexico dealing with risk perceptions and acceptance related to the Waste Isolation Pilot Project (WIPP), a permanent storage site for radioactive waste located near Carlsbad, New Mexico. WIPP became operational and received its first shipment of waste on March 26, 1999. This study tracks the changes of risk perception and acceptance over a decade, using measures taken from 35 statewide surveys of New Mexico citizens spanning the 11-year period from fall 1990 to summer 2001. This time span includes periods before and after WIPP became operational. We find that acceptance of WIPP is greater among those in the most proximate counties to WIPP. Surprisingly, and contrary to expectations drawn from the broader literature, acceptance is also greater among those who live closest to the nuclear waste transportation route. We also find that ideology, partisanship, government approval and broader environmental concerns influence support for WIPP acceptance. Finally, the sequence of procedural steps taken toward formal approval of WIPP by government agencies proved to be important to public acceptance, the most significant being the opening of the WIPP facility itself

Pretreatment of Tc-Containing Waste and Its Effect on Tc-99 Leaching From Grouts

International Nuclear Information System (INIS)

Aloy, Albert; Kovarskaya, Elena N.; Harbour, John R.; Langton, Christine A.; Holtzscheiter, E. William

2007-01-01

A salt solution (doped with Tc-99), that simulates the salt waste stream to be processed at the Saltstone Production Facility, was immobilized in grout waste forms with and without (1) ground granulated blast furnace slag and (2) pretreatment with iron salts. The degree of immobilization of Tc-99 was measured through monolithic and crushed grout leaching tests. Although Fe (+2) was shown to be effective in reducing Tc-99 to the +4 state, the strong reducing nature of the blast furnace slag present in the grout formulation dominated the reduction of Tc-99 in the cured grouts. An effective diffusion coefficient of 4.75 x 10 -12 (Leach Index of 11.4) was measured using the ANSI/ANS-16.1 protocol. The leaching results show that, even in the presence of a concentrated salt solution, blast furnace slag can effectively reduce pertechnetate to the immobile +4 oxidation state. The measured diffusivity was introduced into a flow and transport model (PORFLOW) to calculate the release of Tc-99 from a Saltstone Vault as a function of hydraulic conductivity of the matrix. (authors)

Review Of Concrete Biodeterioration In Relation To Buried Nuclear Waste

International Nuclear Information System (INIS)

Turick, C.

2012-01-01

Long-term storage of low level radioactive material in below ground concrete disposal units (DUs) (Saltstone Disposal Facility) is a means of depositing wastes generated from nuclear operations of the U.S. Department of Energy. Based on the currently modeled degradation mechanisms, possible microbial induced effects on the structural integrity of buried low level wastes must be addressed. Previous international efforts related to microbial impacts on concrete structures that house low level radioactive waste showed that microbial activity can play a significant role in the process of concrete degradation and ultimately structural deterioration. This literature review examines the recent research in this field and is focused on specific parameters that are applicable to modeling and prediction of the fate of concrete vaults housing stored wastes and the wastes themselves. Rates of concrete biodegradation vary with the environmental conditions, illustrating a need to understand the bioavailability of key compounds involved in microbial activity. Specific parameters require pH and osmotic pressure to be within a certain range to allow for microbial growth as well as the availability and abundance of energy sources like components involved in sulfur, iron and nitrogen oxidation. Carbon flow and availability are also factors to consider in predicting concrete biodegradation. The results of this review suggest that microbial activity in Saltstone, (grouted low level radioactive waste) is unlikely due to very high pH and osmotic pressure. Biodegradation of the concrete vaults housing the radioactive waste however, is a possibility. The rate and degree of concrete biodegradation is dependent on numerous physical, chemical and biological parameters. Results from this review point to parameters to focus on for modeling activities and also, possible options for mitigation that would minimize concrete biodegradation. In addition, key chemical components that drive microbial

Review of Concrete Biodeterioration in Relation to Buried Nuclear Waste

Energy Technology Data Exchange (ETDEWEB)

Turick, C; Berry, C.

2012-10-15

Long-term storage of low level radioactive material in below ground concrete disposal units (DUs) (Saltstone Disposal Facility) is a means of depositing wastes generated from nuclear operations of the U.S. Department of Energy. Based on the currently modeled degradation mechanisms, possible microbial induced effects on the structural integrity of buried low level wastes must be addressed. Previous international efforts related to microbial impacts on concrete structures that house low level radioactive waste showed that microbial activity can play a significant role in the process of concrete degradation and ultimately structural deterioration. This literature review examines the recent research in this field and is focused on specific parameters that are applicable to modeling and prediction of the fate of concrete vaults housing stored wastes and the wastes themselves. Rates of concrete biodegradation vary with the environmental conditions, illustrating a need to understand the bioavailability of key compounds involved in microbial activity. Specific parameters require pH and osmotic pressure to be within a certain range to allow for microbial growth as well as the availability and abundance of energy sources like components involved in sulfur, iron and nitrogen oxidation. Carbon flow and availability are also factors to consider in predicting concrete biodegradation. The results of this review suggest that microbial activity in Saltstone, (grouted low level radioactive waste) is unlikely due to very high pH and osmotic pressure. Biodegradation of the concrete vaults housing the radioactive waste however, is a possibility. The rate and degree of concrete biodegradation is dependent on numerous physical, chemical and biological parameters. Results from this review point to parameters to focus on for modeling activities and also, possible options for mitigation that would minimize concrete biodegradation. In addition, key chemical components that drive microbial

Mechanisms of contaminant migration from grouted waste

International Nuclear Information System (INIS)

Magnuson, S.O.; Yu, A.D.

1992-01-01

Low-level radioactive decontaminated salt solution is generated at the Savannah River Site (SRS) from the In-Tank Precipitation process. The solution is mixed with cement, slag, and fly ash, to form a grout, termed ''Saltstone'', that will be disposed in concrete vaults at the Saltstone Disposal Facility (SDF) [1]. Of the contaminants in the Saltstone, the greatest concern to SRS is the potential release of nitrate to the groundwater because of the high initial nitrate concentration (0.25 g/cm 3 ) in the Saltstone and the low Safe Drinking Water Act (SDWA) maximum contaminant level (MCL) of 44 mg/L. The SDF is designed to allow a slow, controlled release over thousands of years. This paper addresses a modeling study of nitrate migration from intact non-degraded concrete vaults in the unsaturated zone for the Radiological Performance Assessment (PA) of the SRS Saltstone Disposal Facility [3]. The PA addresses the performance requirements mandated by DOE Order 5820.2A [4

Acceptance test plan for the Waste Information Control System

International Nuclear Information System (INIS)

Flynn, D.F.

1994-01-01

This document describes the acceptance test plan for the WICS system. The Westinghouse Hanford Company (WHC) Hazardous Material Control Group (HMC) of the 222-S Laboratory has requested the development of a system to help resolve many of the difficulties associated with tracking and data collection of containers and drums of waste. This system has been identified as Waste Information and Control System (WICS). The request for developing and implementing WICS has been made to the Automation and Simulation Engineering Group (ASE)

Does the choice of reactor affect public acceptance of wastes?

International Nuclear Information System (INIS)

Inhaber, H.

1993-01-01

A prime goal of this conference is to suggest future reactor types that would produce greater public acceptability. Presumably the wastes generated by these cycles would, because of lesser amounts or activities, engender fewer disputes over policy than in the past. However, the world-wide arguments over low-level wastes (LLW) suggest this intent is not likely to be achieved. While the activity of these wastes is a tiny fraction of high-level wastes (HLW), the controversies over the former, in Korea, the US and elsewhere, have been as great as for the latter. There is no linear relationship between activity and political desirability. What is needed is a new approach to disposing of and siting all nuclear wastes: LLW, mixed and HLW

A comparison and cross-reference of commercial low-level radioactive waste acceptance criteria

International Nuclear Information System (INIS)

Kerr, T.A.

1997-04-01

This document, prepared by the National Low-Level Waste Management Program at the Idaho National Engineering and Environmental Laboratory, is a comparison and cross-reference of commercial low-level radioactive waste acceptance criteria. Many of these are draft or preliminary criteria as well as implemented criteria at operating low-level radioactive waste management facilities. Waste acceptance criteria from the following entities are included: US Nuclear Regulatory Commission, South Carolina, Washington, Utah, Nevada, California, illinois, Texas, North Carolina, Nebraska, Pennsylvania, New York, and the Midwest Compact Region. Criteria in the matrix include the following: physical form, chemical form, liquid limits, void space in packages, concentration averaging, types of packaging, chelating agents, solidification media, stability requirements, sorptive media, gas, oil, biological waste, pyrophorics, source material, special nuclear material, package dimensions, incinerator ash, dewatered resin, transuranics, and mixed waste. Each criterion in the matrix is cross-referenced to its source document so that exact requirements can be determined

Development of acceptance specifications for low-activity waste from the Hanford tanks

International Nuclear Information System (INIS)

Cunnane, J.C.; Kier, P.H.; Brown, N.R.

1997-01-01

Low-activity products will be in the form of soldified waste and optional matrix and filler materials enclosed in sealed metal boxes. Acceptance specifications limit the physical characteristics of the containers, the chemical and physical characteristics of the waste form and other materials that may be in the container, the waste loading, and the radionuclide leaching characteristics of the waste form. The specifications are designed to ensure that low-activity waste products will be compatible with the driving regulatory and operational requirements and with existing production technologies

Analysis of local acceptance of a radioactive waste disposal facility.

Science.gov (United States)

Chung, Ji Bum; Kim, Hong-Kew; Rho, Sam Kew

2008-08-01

Like many other countries in the world, Korea has struggled to site a facility for radioactive waste for almost 30 years because of the strong opposition from local residents. Finally, in 2005, Gyeongju was established as the first Korean site for a radioactive waste facility. The objectives of this research are to verify Gyeongju citizens' average level of risk perception of a radioactive waste disposal facility as compared to other risks, and to explore the best model for predicting respondents' acceptance level using variables related to cost-benefit, risk perception, and political process. For this purpose, a survey is conducted among Gyeongju residents, the results of which are as follows. First, the local residents' risk perception of an accident in a radioactive waste disposal facility is ranked seventh among a total of 13 risks, which implies that nuclear-related risk is not perceived very highly by Gyeongju residents; however, its characteristics are still somewhat negative. Second, the comparative regression analyses show that the cost-benefit and political process models are more suitable for explaining the respondents' level of acceptance than the risk perception model. This may be the result of the current economic depression in Gyeongju, residents' familiarity with the nuclear industry, or cultural characteristics of risk tolerance.

HWVP compliance with the Hanford site solid waste acceptance criteria

International Nuclear Information System (INIS)

Bromm, R.; Ornelas, J.; Fundingsland, S.; Shah, K.

1993-01-01

In order to ensure that the Hanford Waste Vitrification Project (HWVP) will meet solid waste acceptance criteria, a review of the criteria was performed. The primary purpose of the study was to evaluate the modifications that will be required to bring the HWVP into compliance for secondary waste which will be generated during normal operations of the facility. To accomplish this objective, the current HWVP design was evaluated based on the criteria established. Once the non-compliance areas and potentially non-compliance areas were identified, alternative plant design modifications were proposed. This paper summarizes the results and recommendations of that study

Risk perception, risk evaluation and human values: cognitive bases of acceptability of a radioactive waste repository

International Nuclear Information System (INIS)

Earle, T.C.; Lindell, M.K.; Rankin, W.L.

1981-07-01

Public acceptance of radioactive waste management alternatives depends in part on public perception of the associated risks. Three aspects of those perceived risks were explored in this study: (1) synthetic measures of risk perception based on judgments of probability and consequences; (2) acceptability of hypothetical radioactive waste policies, and (3) effects of human values on risk perception. Both the work on synthetic measures of risk perception and on the acceptability of hypothetical policies included investigations of three categories of risk: (1) Short-term public risk (affecting persons living when the wastes are created), (2) Long-term public risk (affecting persons living after the time the wastes were created), and (3) Occupational risk (affecting persons working with the radioactive wastes). The human values work related to public risk perception in general, across categories of persons affected

Degradation of cementitious materials associated with salstone disposal units

Energy Technology Data Exchange (ETDEWEB)

Flach, G. P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Smith, F. G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2014-09-01

The Saltstone facilities at the DOE Savannah River Site (SRS) stabilize and dispose of low-level radioactive salt solution originating from liquid waste storage tanks at the site. The Saltstone Production Facility (SPF) receives treated salt solution and mixes the aqueous waste with dry cement, blast furnace slag, and fly ash to form a grout slurry which is mechanically pumped into concrete disposal cells that compose the Saltstone Disposal Facility (SDF). The solidified grout is termed “saltstone”. Cementitious materials play a prominent role in the design and long-term performance of the SDF. The saltstone grout exhibits low permeability and diffusivity, and thus represents a physical barrier to waste release. The waste form is also reducing, which creates a chemical barrier to waste release for certain key radionuclides, notably Tc-99. Similarly, the concrete shell of a saltstone disposal unit (SDU) represents an additional physical and chemical barrier to radionuclide release to the environment. Together the waste form and the SDU compose a robust containment structure at the time of facility closure. However, the physical and chemical state of cementitious materials will evolve over time through a variety of phenomena, leading to degraded barrier performance over Performance Assessment (PA) timescales of thousands to tens of thousands of years. Previous studies of cementitious material degradation in the context of low-level waste disposal have identified sulfate attack, carbonation influenced steel corrosion, and decalcification (primary constituent leaching) as the primary chemical degradation phenomena of most relevance to SRS exposure conditions. In this study, degradation time scales for each of these three degradation phenomena are estimated for saltstone and concrete associated with each SDU type under conservative, nominal, and best estimate assumptions.

Identification of items and activities important to waste form acceptance by Westinghouse GoCo sites

International Nuclear Information System (INIS)

Plodinec, M.J.; Marra, S.L.; Dempster, J.; Randklev, E.H.

1993-01-01

The Department of Energy has established specifications (Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms, or WAPS) for canistered waste forms produced at Hanford, Savannah River, and West Valley. Compliance with these specifications requires that each waste form producer identify the items and activities which must be controlled to ensure compliance. As part of quality assurance oversight activities, reviewers have tried to compare the methodologies used by the waste form producers to identify items and activities important to waste form acceptance. Due to the lack of a documented comparison of the methods used by each producer, confusion has resulted over whether the methods being used are consistent. This confusion has been exacerbated by different systems of nomenclature used by each producer, and the different stages of development of each project. The waste form producers have met three times in the last two years, most recently on June 28, 1993, to exchange information on each producer's program. These meetings have been sponsored by the Westinghouse GoCo HLW Vitrification Committee. This document is the result of this most recent exchange. It fills the need for a documented comparison of the methodologies used to identify items and activities important to waste form acceptance. In this document, the methodology being used by each waste form producer is summarized, and the degree of consistency among the waste form producers is determined

Acceptance issues for large items and difficult waste

International Nuclear Information System (INIS)

Palmer, J.; Lock, Peter

2002-01-01

Peter Lock described some particular cases which had given rise to difficult acceptance issues at NIREX, ranging from large size items to the impacts of chemicals used during decontamination on the mobility of radionuclides in a disposal facility: The UK strategy for intermediate level and certain low level radioactive waste disposal is based on production of cementitious waste-forms packaged in a standard range of containers as follows: 500 litre Drum - the normal container for most operational ILW (0.8 m diameter x 1.2 m high); 3 m"3 Box - a larger container for solid wastes (1.72 m x 1.72 m plan x 1.2 m high); 3 m"3 Drum - a larger container for in-drum mixing and immobilisation of sludge waste-forms (1.72 m diameter x 1.2 m high); 4 m Box - for large items of waste, especially from decommissioning (4.0 m x 2.4 m plan x 2.2 m high); 2 m LLW Box - for higher-density wastes (2.0 m x 2.4 m plan x 2.2 m high). In addition the majority of LLW is packaged by supercompaction followed by grouting in modified ISO freight containers (6 m x 2.5 m x 2.5 m). Some wastes do not fit easily into this strategy. These wastes include: very large items, (too big for the 4 m box) which, if dealt with whole, pose transport and disposal problems. These items are discussed further in Section 2; waste whose characteristics make packaging difficult. Such wastes are described in more detail in Section 3

Technology, socio-political acceptance, and the low-level radioactive waste disposal facility

International Nuclear Information System (INIS)

Andrews, L.J.; Domenech, J.S.

1986-01-01

The technology which is required to develop and operate low-level radioactive waste disposal sites in the 1990's is available today. The push for best available technology is a response to the political difficulties in securing public acceptance of the site selection process. Advances in waste management technologies include development of High Integrity Containers (HIC), solidification media, liquid volume reduction techniques using GEODE/sub sm/ and DeVoe-Holbein technology of selective removal of target radioisotopes, and CASTOR V storage casks. Advances in technology alone, however, do not make the site selection process easier and without socio-political acceptance there may be no process at all. Chem-Nuclear has been successful in achieving community acceptance at the Barnwell facility and elsewhere. For example, last June in Fall River County, South Dakota, citizens voted almost 2:1 to support the development of a low-level radioactive waste disposal facility. In Edgemont, the city nearest the proposed site, 85% of the voters were in favor of the proposed facility

Classification of radioactive waste and determination of waste specifications as well as conditions of acceptance for ultimate storage

International Nuclear Information System (INIS)

Merz, E.

1983-04-01

The determination of waste specification and conditions of acceptance must follow a certain scheme, the basics of which will be presented. First the types of waste and the ultimate storage facilities will be characterized. The various categories of waste will be listed in a universally valid system, and the preliminary conditioning options will be determined. Based on the results of safety analysis taking into account the whole system - geological circumstances, ultimate store mines, types and forms of waste - specifications for the various ultimate store products are to be derived following iterative methods. Suggestions though not of a binding nature and probably subject to eventual revisions in part will be presented. To ensure the safety goals, i.e. the exclusion of radioactivity from the human biosphere, appropriate quality control is required concerning the production and the acceptance at the ultimate store. The guiding principles to be heeded will be discussed in brief. (orig./HP) [de

Qualitative acceptance criteria for radioactive wastes to be disposed of in deep geological formations

International Nuclear Information System (INIS)

1990-05-01

The present Safety Guide has to be seen as a companion document to the IAEA Safety Series No. 99. It is concerned with the waste form which is an important component of the overall disposal system. Because of the broad range of waste types and conditioned forms and variations in the sites, designs and constructional approaches being considered for deep geological repositories, this report necessarily approaches the waste acceptance criteria in a general way, recognizing that the assignment of quantitative limits to these criteria has to be the responsibility of national authorities. The main objective of this Safety Guide is to set out qualitative waste acceptance criteria as a basis for specifying quantitative limits for the waste forms and packages which are intended to be disposed of in deep geological repositories. It should serve as guidance for assigning such parameter values which would fully comply with the safety assessment and performance of a waste disposal system as a whole. This document is intended to serve both national authorities and regulatory bodies involved in the development of deep underground disposal systems. The qualitative waste acceptance criteria dealt with in the present Safety Guide are primarily concerned with the disposal of high level, intermediate level and long-lived alpha bearing wastes in deep geological repositories. Although some criteria are also applicable in other waste disposal concepts, it has to be borne in mind that the set of criteria presented here shall ensure the isolation capability of a waste disposal system for periods of time much longer than for other waste streams with shorter lifetimes. 51 refs, 1 tab

Results for the first quarter calendar year 2017 tank 50H salt solution sample

Energy Technology Data Exchange (ETDEWEB)

Crawford, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-04-12

In this memorandum, the chemical and radionuclide contaminant results from the First Quarter Calendar Year 2017 (CY17) sample of Tank 50H salt solution are presented in tabulated form. The First Quarter CY17 Tank 50H samples [a 200 mL sample obtained 6” below the surface (HTF-50-17-7) and a 1 L sample obtained 66” from the tank bottom (HTF-50-17-8)] were obtained on January 15, 2017 and received at Savannah River National Laboratory (SRNL) on January 16, 2017. Prior to obtaining the samples from Tank 50H, a single pump was run at least 4.4 hours and the samples were pulled immediately after pump shut down. All volatile organic analysis (VOA) and semi-volatile organic analysis (SVOA) were performed on the surface sample and all other analyses were performed on the variable depth sample. The information from this characterization will be used by Savannah River Remediation (SRR) for the transfer of aqueous waste from Tank 50H to the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. The chemical and radionuclide contaminant results from the characterization of the First Quarter CY17 sampling of Tank 50H were requested by SRR personnel and details of the testing are presented in the SRNL Task Technical and Quality Assurance Plan (TTQAP). This memorandum is part of Deliverable 2 from SRR request. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the TTQAP for the Tank 50H saltstone task.

Public perception and acceptance of the siting of nuclear waste facilities in seven countries

International Nuclear Information System (INIS)

Numark, N.J.; Paige, H.W.; Wonder, E.F.

1989-09-01

This report was prepared by ERC Environmental and Energy Services Co. (ERCE) on behalf of the Pacific Northwest Laboratory (PNL) and the US Department of Energy (DOE) between February and August 1989. It updates previous reports prepared by ERCE on public acceptance of waste management activities in foreign countries. The report is intended to serve as an aid in understanding experiences with public acceptance of waste activities in foreign countries, and thereby benefit US efforts with respect to public acceptance based on lessons learned abroad. Seven countries are addressed in the report: Belgium, the Federal Republic of Germany, France, Japan, Sweden, Switzerland, and the United Kingdom. The information provided in this report was obtained both from direct interviews of the responsible waste management officials in the seven countries surveyed and from source documents provided by these individuals

Risk perception, risk evaluation and human values: cognitive bases of acceptability of a radioactive waste repository

Energy Technology Data Exchange (ETDEWEB)

Earle, T.C.; Lindell, M.K.; Rankin, W.L.

1981-07-01

Public acceptance of radioactive waste management alternatives depends in part on public perception of the associated risks. Three aspects of those perceived risks were explored in this study: (1) synthetic measures of risk perception based on judgments of probability and consequences; (2) acceptability of hypothetical radioactive waste policies, and (3) effects of human values on risk perception. Both the work on synthetic measures of risk perception and on the acceptability of hypothetical policies included investigations of three categories of risk: (1) Short-term public risk (affecting persons living when the wastes are created), (2) Long-term public risk (affecting persons living after the time the wastes were created), and (3) Occupational risk (affecting persons working with the radioactive wastes). The human values work related to public risk perception in general, across categories of persons affected. Respondents were selected according to a purposive sampling strategy.

Reversing nuclear opposition: evolving public acceptance of a permanent nuclear waste disposal facility.

Science.gov (United States)

Jenkins-Smith, Hank C; Silva, Carol L; Nowlin, Matthew C; deLozier, Grant

2011-04-01

Nuclear facilities have long been seen as the top of the list of locally unwanted land uses (LULUs), with nuclear waste repositories generating the greatest opposition. Focusing on the case of the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, we test competing hypotheses concerning the sources of opposition and support for siting the facility, including demographics, proximity, political ideology, and partisanship, and the unfolding policy process over time. This study tracks the changes of risk perception and acceptance of WIPP over a decade, using measures taken from 35 statewide surveys of New Mexico citizens spanning an 11-year period from fall 1990 to summer 2001. This time span includes periods before and after WIPP became operational. We find that acceptance of WIPP is greater among those whose residences are closest to the WIPP facility. Surprisingly, and contrary to expectations drawn from the broader literature, acceptance is also greater among those who live closest to the nuclear waste transportation route. We also find that ideology, partisanship, government approval, and broader environmental concerns influence support for WIPP acceptance. Finally, the sequence of procedural steps taken toward formal approval of WIPP by government agencies proved to be important to gaining public acceptance, the most significant being the opening of the WIPP facility itself. © 2010 Society for Risk Analysis.

Measurement methodology for fulfilling of waste acceptance criteria for low and intermediate level radioactive waste in storages - 59016

International Nuclear Information System (INIS)

Sokcic-Kostic, M.; Langer, F.; Schultheis, R.

2012-01-01

Low and intermediate level radioactive waste must be sorted and treated before it can be sent to radioactive waste storage. The waste must fulfil an extensive amount of acceptance criteria (WAC) to guarantee a safe storage period. NUKEM Technologies has a broad experience with the building and management of radioactive waste treatment facilities and has developed methods and equipment to produce the waste packages and to gather all the required information. In this article we consider low and intermediate level radioactive waste excluding nuclear fuel material, even fresh fuel with low radiation. Only solid radioactive waste (RAW) will be considered. (Liquid RAW is usually processed and solidified before storage. Exception is the reprocessing of nuclear fuel.) Low and intermediate level radioactive waste has to be kept in storage facilities until isotopes are decayed sufficiently and the waste can be released. The storage has to fulfil certain conditions regarding the possible radiological impact and the possible chemical impact on the environment. With the inventory of nuclear waste characterised, the radiological impact can be estimated. RAW mainly originates from the operation of nuclear power plants. A small amount comes from reprocessing installations or from research entities. Chemical safety aspects are of qualitative nature, excluding substances in whole but not compared to limit values. Therefore they have minor influence on the storage conditions. Hereby corrosion and immobilisation of the waste play important roles. The storage concept assumes that the waste will be released if the radioactivity has decreased to an acceptable level. NUKEM Technologies has been specialised on collecting all data needed for the fulfilling of waste acceptance criteria (WAC). The classification as low or intermediate level waste is made on base of surface dose rate of the waste package as well as on the mass specific beta activity. Low level waste must not include isotopes

Notice of inquiry on waste acceptance issues: Response summary

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-03-01

On May 25, 1994, the Department of Energy published a Notice of Inquiry on Waste Acceptance Issues in the Federal Register. Through this Notice of Inquiry, the Department sought to implement the Secretary`s initiative to explore with affected parties various options and methods for sharing the costs related to the financial burden associated with continued on-site storage by eliciting the views of affected parties on: (1) The Department`s preliminary view that it does not have a statutory obligation to begin accepting spent nuclear fuel in 1998 in the absence of an operational repository or other suitable storage facility constructed under the Nuclear Waste Policy Act of 1982, as amended; (2) The need for an interim, away-from-reactor storage facility prior to repository operations; and (3) Options for offsetting, through the Nuclear Waste Fund, a portion of the financial burden that may be incurred by utilities in continuing to store spent nuclear fuel at reactor sites beyond 1998. The Department received a total of 1,111 responses representing 1,476 signatories to this Notice of Inquiry. The responses included submittals from utilities (38 responses); public utility/service commissions and utility regulators (26 responses); Federal, state, and local governments, agencies, and representatives (23 responses); industry and companies (30 responses); public interest groups and other organizations (19 responses); and members of the general public (975 responses).

Notice of inquiry on waste acceptance issues: Response summary

International Nuclear Information System (INIS)

1995-03-01

On May 25, 1994, the Department of Energy published a Notice of Inquiry on Waste Acceptance Issues in the Federal Register. Through this Notice of Inquiry, the Department sought to implement the Secretary's initiative to explore with affected parties various options and methods for sharing the costs related to the financial burden associated with continued on-site storage by eliciting the views of affected parties on: (1) The Department's preliminary view that it does not have a statutory obligation to begin accepting spent nuclear fuel in 1998 in the absence of an operational repository or other suitable storage facility constructed under the Nuclear Waste Policy Act of 1982, as amended; (2) The need for an interim, away-from-reactor storage facility prior to repository operations; and (3) Options for offsetting, through the Nuclear Waste Fund, a portion of the financial burden that may be incurred by utilities in continuing to store spent nuclear fuel at reactor sites beyond 1998. The Department received a total of 1,111 responses representing 1,476 signatories to this Notice of Inquiry. The responses included submittals from utilities (38 responses); public utility/service commissions and utility regulators (26 responses); Federal, state, and local governments, agencies, and representatives (23 responses); industry and companies (30 responses); public interest groups and other organizations (19 responses); and members of the general public (975 responses)

Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

International Nuclear Information System (INIS)

1990-03-01

This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs

Acceptable Knowledge Summary Report for Waste Stream: SR-T001-221F-HET/Drums

Energy Technology Data Exchange (ETDEWEB)

Lunsford, G.F.

1998-10-26

Since beginning operations in 1954, the Savannah River Site FB-Line produced Weapons Grade Plutonium for the United States National Defense Program. The facility mission was mainly to process dilute plutonium solution received from the 221-F Canyon into highly purified plutonium metal. As a result of various activities (maintenance, repair, clean up, etc.) in support of the mission, the facility generated a transuranic heterogeneous debris waste stream. Prior to January 25, 1990, the waste stream was considered suspect mixed transuranic waste (based on potential for inclusion of F-Listed solvent rags/wipes) and is not included in this characterization. Beginning January 25, 1990, Savannah River Site began segregation of rags and wipes containing F-Listed solvents thus creating a mixed transuranic waste stream and a non-mixed transuranic waste stream. This characterization addresses the non-mixed transuranic waste stream packaged in 55-gallon drums after January 25, 1990.Characterization of the waste stream was achieved using knowledge of process operations, facility safety basis documentation, facility specific waste management procedures and storage / disposal records. The report is fully responsive to the requirements of Section 4.0 "Acceptable Knowledge" from the WIPP Transuranic Waste Characterization Quality Assurance Plan, CAO-94-1010, and provides a sound, (and auditable) characterization that satisfies the WIPP criteria for Acceptable Knowledge.

Waste acceptance criteria study: Volume 2, Appendixes: Final report

International Nuclear Information System (INIS)

Johnson, E.R.; McLeod, N.B.; McBride, J.A.

1988-09-01

These appendices to the report on Waste Acceptance Criteria have been published as a separate volume for the convenience of the reader. They consist of the text of the 10CFR961 Contract for disposal of spent fuel, estimates of the cost (savings) to the DOE system of accepting different forms of spent fuel, estimates of costs of acceptance testing/inspection of spent fuel, illustrative specifications and procedures, and the resolution of comments received on a preliminary draft of the report. These estimates of costs contained herein preliminary and are intended only to demonstrate the trends in costs, the order of magnitude involved, and the methodology used to develop the costs. The illustrative specifications and procedures included herein have been developed for the purpose of providing a starting point for the development of a consensus on such matters between utilities and DOE

Dissolution test for low-activity waste product acceptance

International Nuclear Information System (INIS)

Ebert, W. L.

1998-01-01

We have measured the mean and standard deviation of the solution concentrations of B, Na, and Si attained in replicate dissolution tests conducted at temperatures of 20, 40, and 70 C, for durations of 3 and 7 days, and at glass/water mass ratios of 1:10 and 1:1. These and other tests were conducted to evaluate the adequacy of the test methods specified in privatization contracts and to develop a data base that can be used to evaluate the reliability of reported results for tests performed on the waste products. Tests were conducted with a glass that we formulated to be similar to low-activity waste products that will be produced during the remediation of Hanford tank wastes. Statistical analyses indicated that, while the mean concentrations of B, Na, and Si were affected by the values of test parameters, the standard deviation of replicate tests was not. The precision of the tests was determined primarily by uncertainties in the analysis of the test solutions. Replicate measurements of other glass properties that must be reported for Hanford low-activity waste products were measured to evaluate the possible adoption of the glass used in these tests as a standard test material for the product acceptance process

TRU [transuranic] waste certification compliance requirements for acceptance of newly generated contact-handled wastes to be shipped to the Waste Isolation Pilot Plant: Revision 2

International Nuclear Information System (INIS)

1989-01-01

Compliance requirements are presented for certifying that unclassified, newly generated (NG), contact-handled (CH) transuranic (TRU) solid wastes from defense programs meet the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC). Where appropriate, transportation and interim storage requirements are incorporated; however, interim storage sites may have additional requirements consistent with these requirements. All applicable Department of Energy (DOE) orders must continue to be met. The compliance requirements for stored or buried waste are not addressed in this document. The compliance requirements are divided into four sections, primarily determined by the general feature that the requirements address. These sections are General Requirements, Waste Container Requirements, Waste Form Requirements, and Waste Package Requirements. The waste package is the combination of waste container and waste. 10 refs., 1 fig

Acceptable Knowledge Summary Report for Waste Stream: SR-T001-221F-HET/Drums

International Nuclear Information System (INIS)

Lunsford, G.F.

1999-01-01

This report is fully responsive to the requirements of Section 4.0 ''Acceptable Knowledge'' from the WIPP Transuranic Waste Characterization Quality Assurance Plan, CAO-94-1010, and provides a sound, (and auditable) characterization that satisfies the WIPP criteria for Acceptable Knowledge

Public acceptance of radioactive waste transportation systems

International Nuclear Information System (INIS)

Gablin, K.A.

1978-01-01

As the thoughts of the country concentrate on the problems of transportation of waste through high traffic urban areas, the problem of how to deal directly and honestly with the public takes on greater significance in the nuclear industry. Non-technical aspects of the methods of transportation, especially by railroad and highway, enter into the total scheme of moving radioactive waste from both weapon and nuclear power plant sources to final processing and disposal. Factors such as shape, color, size, familiarity, and industrial designing are necessary ingredients that take on equal or more significance that the designing of containers to survive the hypothetical accident conditions of the present, or even of the future. Protective Packaging, Inc. has been a leader in the presentation of containers to the private and public sector of the nuclear industry. The products have undergone very open testing, in public, with both invited and uninvited witnesses. In those experiences, dating back to 1969, the problems of public acceptance will be related between the technical problems and the associated social and political problems that relate to container acceptance by the public in today's world. Proven experience data, relative to the safety of the present day systems will be discussed, as well as methods of improving the image in the future. Review will also be given to the effort by industry to discuss the proven record with parties outside the nuclear industry, i.e., individuals and pressure groups that are diametrically opposed to review the facts relative to safety as opposed to other, but more traditional industries

Acceptable knowledge summary report for combustible/noncombustible, metallic, and HEPA filter waste resulting from 238Pu fabrication activities

International Nuclear Information System (INIS)

Rogers, P.S.Z.; Foxx, C.L.

1998-01-01

All transuranic (TRU) waste must be sufficiently characterized and certified before it is shipped to the Waste Isolation Pilot Plant (WIPP). The US Environmental Protection Agency (EPA) allows use of acceptable knowledge (AK) for waste characterization. EPA uses the term AK in its guidance document and defines AK and provides guidelines on how acceptable knowledge should be obtained and documented. This AK package has been prepared in accordance with Acceptable Knowledge Documentation (TWCP-QP-1.1-021,R.2). This report covers acceptable knowledge information for five waste streams generated at TA-55 during operations to fabricate various heat sources using feedstock 238 Pu supplied by the Savannah River Site (SRS). The 238 Pu feedstock itself does not contain quantities of RCRA-regulated constituents above regulatory threshold limits, as known from process knowledge at SRS and as confirmed by chemical analysis. No RCRA-regulated chemicals were used during 238 Pu fabrication activities at TA-55, and all 238 Pu activities were physically separated from other plutonium processing activities. Most of the waste generated from the 238 Pu fabrication activities is thus nonmixed waste, including waste streams TA-55-43, 45, and 47. The exceptions are waste streams TA-55-44, which contains discarded lead-lined rubber gloves used in the gloveboxes that contained the 238 Pu material, and TA-55-46, which may contain pieces of discarded lead. These waste streams have been denoted as mixed because of the presence of the lead-containing material

Development of Waste Acceptance Criteria at 221-U Building: Initial Flow and Transport Scoping Calculations

Energy Technology Data Exchange (ETDEWEB)

Freedman, Vicky L.; Zhang, Z. F.; Keller, Jason M.; Chen, Yousu

2007-05-30

This report documents numerical flow and transport simulations performed that establish initial waste acceptance criteria for the potential waste streams that may be safely sequestered in the 221-U Building and similar canyon structures. Specifically, simulations were executed to identify the maximum loading of contaminant mass (without respect to volume) that can be emplaced within the 221-U Building with no more than 1 pCi/m2 of contaminant migrating outside the structure within a 1,000 year time period. The initial scoping simulations were executed in one dimension to assess important processes, and then two dimensions to establish waste acceptance criteria. Two monolithic conditions were assessed: (1) a grouted canyon monolith; and (2) a canyon monolith filled with sand, both assuming no cracks or fissures were present to cause preferential transport. A three-staged approach was taken to account for different processes that may impact the amount of contaminant that can be safely sequestered in canyon structure. In the first stage, flow and transport simulations established waste acceptance criteria based on a linear (Kd) isotherm approach. In the second stage, impacts on thermal loading were examined and the differences in waste acceptance criteria quantified. In the third stage of modeling, precipitation/dissolution reactions were considered on the release and transport of the contaminants, and the subsequent impact on the maximum contaminant loading. The reactive transport modeling is considered a demonstration of the reactive transport capability, and shows the importance of its use for future performance predictions once site-specific data have been obtained.

Modeling Analysis For Grout Hopper Waste Tank

International Nuclear Information System (INIS)

Lee, S.

2012-01-01

The Saltstone facility at Savannah River Site (SRS) has a grout hopper tank to provide agitator stirring of the Saltstone feed materials. The tank has about 300 gallon capacity to provide a larger working volume for the grout nuclear waste slurry to be held in case of a process upset, and it is equipped with a mechanical agitator, which is intended to keep the grout in motion and agitated so that it won't start to set up. The primary objective of the work was to evaluate the flow performance for mechanical agitators to prevent vortex pull-through for an adequate stirring of the feed materials and to estimate an agitator speed which provides acceptable flow performance with a 45 o pitched four-blade agitator. In addition, the power consumption required for the agitator operation was estimated. The modeling calculations were performed by taking two steps of the Computational Fluid Dynamics (CFD) modeling approach. As a first step, a simple single-stage agitator model with 45 o pitched propeller blades was developed for the initial scoping analysis of the flow pattern behaviors for a range of different operating conditions. Based on the initial phase-1 results, the phase-2 model with a two-stage agitator was developed for the final performance evaluations. A series of sensitivity calculations for different designs of agitators and operating conditions have been performed to investigate the impact of key parameters on the grout hydraulic performance in a 300-gallon hopper tank. For the analysis, viscous shear was modeled by using the Bingham plastic approximation. Steady state analyses with a two-equation turbulence model were performed. All analyses were based on three-dimensional results. Recommended operational guidance was developed by using the basic concept that local shear rate profiles and flow patterns can be used as a measure of hydraulic performance and spatial stirring. Flow patterns were estimated by a Lagrangian integration technique along the flow paths

Thermal properties of fly ash substituted slag cement waste forms for disposal of Savannah River Plant salt waste

International Nuclear Information System (INIS)

Roy, D.M.; Kaushal, S.; Licastro, P.H.; Langton, C.A.

1985-01-01

Waste processing at the Savannah River Plant will involve reconstitution of the salts (NaNO 3 , NaNO 2 , NaOH, etc.) into a concentrated solution (32 weight percent salts) followed by solidification in a cement-based waste form for burial. The stability and mechanical durability of such a 'saltstone monolith' will depend largely on the temperature reached due to heat of hydration and the thermal properties of the waste form. Fly ash has been used as an inexpensive constituent and to moderate the hydration and setting processes so as to avoid reaching prohibitively high temperatures which could cause thermal stresses. Both high-calcium and low-calcium fly ashes have been studied for this purpose. Other constituents of these mixes include granulated blast furnace slag and finely crushed limestone. Adiabatic temperature increase and thermal conductivity of these mixes have been studied and related x-ray diffraction and scanning electron microscopy studies carried out to understand the hydration process

Review of issues relevant to acceptable risk criteria for nuclear waste management

International Nuclear Information System (INIS)

Cohen, J.J.

1978-01-01

Development of acceptable risk criteria for nuclear waste management requires the translation of publicly determined goals and objectives into definitive issues which, in turn, require resolution. Since these issues are largely of a subjective nature, they cannot be resolved by technological methods. Development of acceptable risk criteria might best be accomplished by application of a systematic methodology for the optimal implementation of subjective values. Multi-attribute decision analysis is well suited for this purpose

The role of waste package specifications as a forerunner to ILW repository conditions for acceptance

International Nuclear Information System (INIS)

Barlow, S.V.; Palmer, J.D.

1998-01-01

In the absence of a finalized repository site, design or associated safety case, Nirex is not in a position to issue conditions for acceptance. Nirex has therefore developed a strategy which facilitates packaging of intermediate level waste by providing guidance through waste package specifications, supported by the formal assessment of specific packaging proposals on a case-by-case basis. The waste package specifications are comprehensive and cover all aspects of the waste package including dimensions and other key features, performance standards, wasteform, quality assurance, and data recording requirements. The waste package specifications will be subject to periodic review as repository design and safety cases are finalized and will progressively become site- and design-specific. The waste package specifications will eventually form the basis for conditions for acceptance. The strategy described in this paper has been successfully followed by Nirex and customers for the past ten years and has permitted wastes to be packaged for a deep repository with confidence in the absence of a finalized site and safety cases for the repository. Because the process has its basis in a generic repository concept, it remains robust, despite the increased uncertainty following the March 1997 Secretary of State's decision, as to the siting and time-scale of a deep waste repository, and continues to be an important component of the UK's waste management strategy. (author)

Sensitivity analysis and benchmarking of the BLT low-level waste source term code

International Nuclear Information System (INIS)

Suen, C.J.; Sullivan, T.M.

1993-07-01

To evaluate the source term for low-level waste disposal, a comprehensive model had been developed and incorporated into a computer code, called BLT (Breach-Leach-Transport) Since the release of the original version, many new features and improvements had also been added to the Leach model of the code. This report consists of two different studies based on the new version of the BLT code: (1) a series of verification/sensitivity tests; and (2) benchmarking of the BLT code using field data. Based on the results of the verification/sensitivity tests, the authors concluded that the new version represents a significant improvement and it is capable of providing more realistic simulations of the leaching process. Benchmarking work was carried out to provide a reasonable level of confidence in the model predictions. In this study, the experimentally measured release curves for nitrate, technetium-99 and tritium from the saltstone lysimeters operated by Savannah River Laboratory were used. The model results are observed to be in general agreement with the experimental data, within the acceptable limits of uncertainty

Waste vitrification: prediction of acceptable compositions in a lime-soda-silica glass-forming system

International Nuclear Information System (INIS)

Gilliam, T.M.; Jantzen, C.M.

1996-10-01

A model is presented based upon calculated bridging oxygens which allows the prediction of the region of acceptable glass compositions for a lime-soda-silica glass-forming system containing mixed waste. The model can be used to guide glass formulation studies (e.g., treatability studies) or assess the applicability of vitrification to candidate waste streams

Performance Assessment Program for the Savannah River Site Liquid Waste Facilities - 13610

Energy Technology Data Exchange (ETDEWEB)

Rosenberger, Kent H. [Savannah River Remediation LLC, Building 705-1C, Aiken, SC 29808 (United States)

2013-07-01

The Liquid Waste facilities at the U.S. Department of Energy's (DOE) Savannah River Site (SRS) are operated by Liquid Waste Operations contractor Savannah River Remediation LLC (SRR). A separate Performance Assessment (PA) is prepared to support disposal operations at the Saltstone Disposal Facility and closure evaluations for the two liquid waste tank farm facilities at SRS, F-Tank Farm and H-Tank Farm. A PA provides the technical basis and results to be used in subsequent documents to demonstrate compliance with the pertinent requirements identified in operations and closure regulatory guidance. The Saltstone Disposal Facility is subject to a State of South Carolina industrial solid waste landfill permit and the tank farms are subject to a state industrial waste water permit. The three Liquid Waste facilities are also subject to a Federal Facility Agreement approved by the State, DOE and the Environmental Protection Agency (EPA). Due to the regulatory structure, a PA is a key technical document reviewed by the DOE, the State of South Carolina and the EPA. As the waste material disposed of in the Saltstone Disposal Facility and the residual material in the closed tank farms is also subject to reclassification prior to closure via a waste determination pursuant to Section 3116 of the Ronald W. Reagan National Defense Authorization Act of Fiscal Year 2005, the U.S. Nuclear Regulatory Commission (NRC) is also a reviewing agency for the PAs. Pursuant to the Act, the NRC also has a continuing role to monitor disposal actions to assess compliance with stated performance objectives. The Liquid Waste PA program at SRS represents a continual process over the life of the disposal and closure operations. When the need for a PA or PA revision is identified, the first step is to develop a conceptual model to best represent the facility conditions. The conceptual model will include physical dimensions of the closed system, both the engineered and natural system, and

Performance Assessment Program for the Savannah River Site Liquid Waste Facilities - 13610

International Nuclear Information System (INIS)

Rosenberger, Kent H.

2013-01-01

The Liquid Waste facilities at the U.S. Department of Energy's (DOE) Savannah River Site (SRS) are operated by Liquid Waste Operations contractor Savannah River Remediation LLC (SRR). A separate Performance Assessment (PA) is prepared to support disposal operations at the Saltstone Disposal Facility and closure evaluations for the two liquid waste tank farm facilities at SRS, F-Tank Farm and H-Tank Farm. A PA provides the technical basis and results to be used in subsequent documents to demonstrate compliance with the pertinent requirements identified in operations and closure regulatory guidance. The Saltstone Disposal Facility is subject to a State of South Carolina industrial solid waste landfill permit and the tank farms are subject to a state industrial waste water permit. The three Liquid Waste facilities are also subject to a Federal Facility Agreement approved by the State, DOE and the Environmental Protection Agency (EPA). Due to the regulatory structure, a PA is a key technical document reviewed by the DOE, the State of South Carolina and the EPA. As the waste material disposed of in the Saltstone Disposal Facility and the residual material in the closed tank farms is also subject to reclassification prior to closure via a waste determination pursuant to Section 3116 of the Ronald W. Reagan National Defense Authorization Act of Fiscal Year 2005, the U.S. Nuclear Regulatory Commission (NRC) is also a reviewing agency for the PAs. Pursuant to the Act, the NRC also has a continuing role to monitor disposal actions to assess compliance with stated performance objectives. The Liquid Waste PA program at SRS represents a continual process over the life of the disposal and closure operations. When the need for a PA or PA revision is identified, the first step is to develop a conceptual model to best represent the facility conditions. The conceptual model will include physical dimensions of the closed system, both the engineered and natural system, and modeling

Acceptability of a low and intermediate level radioactive waste repository

International Nuclear Information System (INIS)

Zeleznik, N.; Polic, M.

2000-01-01

Siting of a radioactive waste repository, even for the waste of low and intermediate level (LILW) radioactivity, presents a great problem in almost every country that produces such waste. The main problem is not a technical one, but socio-psychological, namely the acceptability of this kind of repository. In general, people are opposed to any such kind of facility in their vicinity (NIMBY). In this study we try to establish the factors that influence people's behavior regarding the construction of a radioactive waste repository in their local community, with the use of Ajzen's model of planned behavior. Two different scenarios about the construction of a radioactive waste repository in their community, together with a set of questions were presented to participants from different schools. Data from the survey were analysed by multivariate methods, and a model of relevant behaviour was proposed. From the results it can be seen that different approaches to local community participation in site selection process slightly influence people's attitudes towards the LILW repository, while significant differences in answers were found in the responses which depend on participants' knowledge. Therefore the RAO Agency will further intensify preparation of the relevant communication plan and start with its implementation to support LILW repository site selection process, which will also include educational programme. (author)

Acceptability criteria for final underground disposal of radioactive waste

International Nuclear Information System (INIS)

Sousselier, Y.

1984-01-01

Specialists now generally agree that the underground disposal of suitably immobilized radioactive waste offers a means of attaining the basic objective of ensuring the immediate and long-term protection of man and the environment throughout the requisite period of time and in all foreseeable circumstances. Criteria of a more general as well as a more specific nature are practical means through which this basic protection objective can be reached. These criteria, which need not necessarily be quantified, enable the authorities to gauge the acceptability of a given project and provide those responsible for waste management with a basis for making decisions. In short, these principles constitute the framework of a suitably safety-oriented waste management policy. The more general criteria correspond to the protection objectives established by the national authorities on the basis of principles and recommendations formulated by international organizations, in particular the ICRP and the IAEA. They apply to any underground disposal system considered as a whole. The more specific criteria provide a means of evaluating the degree to which the various components of the disposal system meet the general criteria. They must also take account of the interaction between these components. As the ultimate aim is the overall safety of the disposal system, individual components can be adjusted to compensate for the performance of others with respect to the criteria. This is the approach adopted by the international bodies and national authorities in developing acceptability criteria for the final underground radioactive disposal systems to be used during the operational and post-operational phases respectively. The main criteria are reviewed and an attempt is made to assess the importance of the specific criteria according to the different types of disposal systems. (author)

A proposed risk acceptance criterion for nuclear fuel waste disposal

International Nuclear Information System (INIS)

Mehta, K.

1985-06-01

The need to establish a radiological protection criterion that applies specifically to disposal of high level nuclear fuel wastes arises from the difficulty of applying the present ICRP recommendations. These recommendations apply to situations in which radiological detriment can be actively controlled, while a permanent waste disposal facility is meant to operate without the need for corrective actions. Also, the risks associated with waste disposal depend on events and processes that have various probabilities of occurrence. In these circumstances, it is not suitable to apply standards that are based on a single dose limit as in the present ICRP recommendations, because it will generally be possible to envisage events, perhaps rare, that would lead to doses above any selected limit. To overcome these difficulties, it is proposed to base a criterion for acceptability on a set of dose values and corresponding limiting values of probabilities; this set of values constitutes a risk-limit line. A risk-limit line suitable for waste disposal is proposed that has characteristics consistent with the basic philosophy of the ICRP and UNSCEAR recommendations, and is based on levels on natural background radiation

An investigation into waste charges in Ireland, with emphasis on public acceptability.

Science.gov (United States)

Dunne, Louise; Convery, Frank J; Gallagher, Louise

2008-12-01

There are 34 local authorities in Ireland with legal responsibility to deal with waste arising in their jurisdictions. In 2003 the National government introduced legislation that allows local authorities to recover the costs of waste collection and disposal, and to do so by 'executive function', i.e., not requiring support or agreement by the relevant local political representatives. The year 2005 was set as the date by which implementation of a pay by weight or volume was to be introduced. The local authorities were given autonomy as to how they addressed this challenge, so we have - in theory - 34 potentially different experiences from which to learn. This paper examines the pay-as-you-throw (PAYT) waste system in Ireland as it develops in line with EU and National demands, with a view to assessing economic and environmental efficiency. All local authorities were surveyed and thirteen responded. While this only represents about 38% of the total number, it includes jurisdictions that contribute in total more than 50% of waste arising. Key figures in the policy and business community were also interviewed in order to identify how the charging schemes were implemented, and to what effect. These insights and parallel investigations are used to review the potential for problems regarding public acceptability of environmental taxes and examine the evidence for economic and environmental efficiency, as well as problem areas, using data from each of the responding local authority jurisdictions. Concentrating on the incentives and drivers across households, municipalities and private waste contractors, the variations in charging system, annual charges and landfill charges are compared where information was available. The various jurisdictions are also examined in terms of relative successes and problems encountered in the transition from fixed charge or free waste collection to PAYT systems. The Irish situation is placed in the context of the international literature on

Radioactive waste management: review on clearance levels and acceptance criteria legislation, requirements and standards.

Science.gov (United States)

Maringer, F J; Suráň, J; Kovář, P; Chauvenet, B; Peyres, V; García-Toraño, E; Cozzella, M L; De Felice, P; Vodenik, B; Hult, M; Rosengård, U; Merimaa, M; Szücs, L; Jeffery, C; Dean, J C J; Tymiński, Z; Arnold, D; Hinca, R; Mirescu, G

2013-11-01

In 2011 the joint research project Metrology for Radioactive Waste Management (MetroRWM)(1) of the European Metrology Research Programme (EMRP) started with a total duration of three years. Within this project, new metrological resources for the assessment of radioactive waste, including their calibration with new reference materials traceable to national standards will be developed. This paper gives a review on national, European and international strategies as basis for science-based metrological requirements in clearance and acceptance of radioactive waste. © 2013 Elsevier Ltd. All rights reserved.

Waste Acceptance Testing of Secondary Waste Forms: Cast Stone, Ceramicrete and DuraLith

International Nuclear Information System (INIS)

Mattigod, Shas V.; Westsik, Joseph H.; Chung, Chul-Woo; Lindberg, Michael J.; Parker, Kent E.

2011-01-01

To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions has initiated secondary-waste-form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is conducting tests on four candidate waste forms to evaluate their ability to meet potential waste acceptance criteria for immobilized secondary wastes that would be placed in the IDF. All three waste forms demonstrated compressive strengths above the minimum 3.45 MPa (500 psi) set as a target for cement-based waste forms. Further, none of the waste forms showed any significant degradation in compressive strength after undergoing thermal cycling (30 cycles in a 10 day period) between -40 C and 60 C or water immersion for 90 days. The three leach test methods are intended to measure the diffusion rates of contaminants from the waste forms. Results are reported in terms of diffusion coefficients and a leachability index (LI) calculated based on the diffusion coefficients. A smaller diffusion coefficient and a larger LI are desired. The NRC, in its Waste Form Technical Position (NRC 1991), provides recommendations and guidance regarding methods to demonstrate waste stability for land disposal of radioactive waste. Included is a recommendation to conduct leach tests using the ANS 16.1 method. The resulting leachability index (LI) should be greater than 6.0. For Hanford secondary wastes, the LI > 6.0 criterion applies to sodium leached from the waste form. For technetium and iodine, higher targets of LI > 9 for Tc and LI > 11 for iodine have been set based on early waste-disposal risk and performance assessment analyses. The results of these three leach tests conducted for a total time between 11days (ASTM C1308) to 90 days (ANS 16.1) showed: (1) Technetium diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that

Waste Acceptance Testing of Secondary Waste Forms: Cast Stone, Ceramicrete and DuraLith

Energy Technology Data Exchange (ETDEWEB)

Mattigod, Shas V.; Westsik, Joseph H.; Chung, Chul-Woo; Lindberg, Michael J.; Parker, Kent E.

2011-08-12

To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions has initiated secondary-waste-form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is conducting tests on four candidate waste forms to evaluate their ability to meet potential waste acceptance criteria for immobilized secondary wastes that would be placed in the IDF. All three waste forms demonstrated compressive strengths above the minimum 3.45 MPa (500 psi) set as a target for cement-based waste forms. Further, none of the waste forms showed any significant degradation in compressive strength after undergoing thermal cycling (30 cycles in a 10 day period) between -40 C and 60 C or water immersion for 90 days. The three leach test methods are intended to measure the diffusion rates of contaminants from the waste forms. Results are reported in terms of diffusion coefficients and a leachability index (LI) calculated based on the diffusion coefficients. A smaller diffusion coefficient and a larger LI are desired. The NRC, in its Waste Form Technical Position (NRC 1991), provides recommendations and guidance regarding methods to demonstrate waste stability for land disposal of radioactive waste. Included is a recommendation to conduct leach tests using the ANS 16.1 method. The resulting leachability index (LI) should be greater than 6.0. For Hanford secondary wastes, the LI > 6.0 criterion applies to sodium leached from the waste form. For technetium and iodine, higher targets of LI > 9 for Tc and LI > 11 for iodine have been set based on early waste-disposal risk and performance assessment analyses. The results of these three leach tests conducted for a total time between 11days (ASTM C1308) to 90 days (ANS 16.1) showed: (1) Technetium diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that

Defining waste acceptance criteria for the Hanford Replacement Cross-Site Transfer System

International Nuclear Information System (INIS)

Hudson, J.D.

1996-04-01

This document provides a methodology for defining waste acceptance criteria for the Hanford Replacement Cross-Site Transfer System (RCSTS). This methodology includes characterization, transport analysis, and control. A framework is described for each of these functions. A tool was developed for performing the calculations associated with the transport analysis. This tool, a worksheet that is available in formats acceptable for a variety of PC spreadsheet programs, enables a comparison of the pressure required to transport a given slurry at a rate that particulate suspension is maintained to the pressure drop available from the RCSTS

WRAP low level waste (LLW) glovebox acceptance test report

International Nuclear Information System (INIS)

Leist, K.J.

1998-01-01

In June 28, 1997, the Low Level Waste (LLW) glovebox was tested using glovebox acceptance test procedure 13031A-85. The primary focus of the glovebox acceptance test was to examine control system interlocks, display menus, alarms, and operator messages. Limited mechanical testing involving the drum ports, hoists, drum lifter, compacted drum lifter, drum tipper, transfer car, conveyors, lidder/delidder device and the supercompactor were also conducted. As of November 24, 1997, 2 of the 131 test exceptions that affect the LLW glovebox remain open. These items will be tracked and closed via the WRAP Master Test Exception Database. As part of Test Exception resolution/closure the responsible individual closing the Test Exception performs a retest of the affected item(s) to ensure the identified deficiency is corrected, and, or to test items not previously available to support testing. Test Exceptions are provided as appendices to this report

WRAP low level waste (LLW) glovebox acceptance test report

Energy Technology Data Exchange (ETDEWEB)

Leist, K.J.

1998-02-17

In June 28, 1997, the Low Level Waste (LLW) glovebox was tested using glovebox acceptance test procedure 13031A-85. The primary focus of the glovebox acceptance test was to examine control system interlocks, display menus, alarms, and operator messages. Limited mechanical testing involving the drum ports, hoists, drum lifter, compacted drum lifter, drum tipper, transfer car, conveyors, lidder/delidder device and the supercompactor were also conducted. As of November 24, 1997, 2 of the 131 test exceptions that affect the LLW glovebox remain open. These items will be tracked and closed via the WRAP Master Test Exception Database. As part of Test Exception resolution/closure the responsible individual closing the Test Exception performs a retest of the affected item(s) to ensure the identified deficiency is corrected, and, or to test items not previously available to support testing. Test Exceptions are provided as appendices to this report.

Development of performance assessment methodology for establishment of quantitative acceptance criteria of near-surface radioactive waste disposal

Energy Technology Data Exchange (ETDEWEB)

Kim, C. R.; Lee, E. Y.; Park, J. W.; Chang, G. M.; Park, H. Y.; Yeom, Y. S. [Korea Hydro and Nuclear Power Co., Ltd., Seoul (Korea, Republic of)

2002-03-15

The contents and the scope of this study are as follows : review of state-of-the-art on the establishment of waste acceptance criteria in foreign near-surface radioactive waste disposal facilities, investigation of radiological assessment methodologies and scenarios, investigation of existing models and computer codes used in performance/safety assessment, development of a performance assessment methodology(draft) to derive quantitatively radionuclide acceptance criteria of domestic near-surface disposal facility, preliminary performance/safety assessment in accordance with the developed methodology.

241-AZ-101 Waste Tank Color Video Camera System Shop Acceptance Test Report

Energy Technology Data Exchange (ETDEWEB)

WERRY, S.M.

2000-03-23

This report includes shop acceptance test results. The test was performed prior to installation at tank AZ-101. Both the camera system and camera purge system were originally sought and procured as a part of initial waste retrieval project W-151.

241-AZ-101 Waste Tank Color Video Camera System Shop Acceptance Test Report

International Nuclear Information System (INIS)

WERRY, S.M.

2000-01-01

This report includes shop acceptance test results. The test was performed prior to installation at tank AZ-101. Both the camera system and camera purge system were originally sought and procured as a part of initial waste retrieval project W-151

Nuclear waste and public acceptance. A study about the situation in the Netherlands

International Nuclear Information System (INIS)

Damveld, H.

1999-01-01

The author bases the paper and study on the concept of the risk society coined by the sociologist Ulrich Beck. Very briefly, risk society means that in modern societies the basic conflicts have shifted from the distribution of prosperity to distribution of risks. The author points out five important factors affecting the willingness to accept a risk, which all have negative indication for the case of nuclear energy. 1. Catastrophe effect: The willingness to accept a more probable risk with less serious consequences is greater than the willingness to accept an extremely unlikely risk with serious consequences. Nuclear energy is considered to be a risk with catastrophe effect. 2. Justice factor: This factor too plays a central role in the negative assessment of nuclear energy. It is considered to be unjust to impose risks on future generations. 3. Voluntariness: The acceptance of a risk increases with the voluntariness. But nuclear energy is not considered to be a voluntary risk. 4. Confidence: The extent of confidence is an important factor determing how information on a possible risk will be assessed. In the Netherlands 68 percent of the population does not trust that the government will provide immediate and honest information on nuclear accidents. 5. Avoidability: An inevitable risk is more likely to be accepted than an avoidable risk. If already produced waste is involved and if there is even a chance that the nuclear power station concerned will be shut-down, the population is more likely to accept the final repository than if it were a case of searching for a final repository for waste not yet produced. (orig.) [de

Determination of acceptable risk criteria for nuclear waste management

International Nuclear Information System (INIS)

Cohen, J.J.

1977-01-01

The initial phase of the work performed during FY 1977 consisted of performing a ''scoping'' study to define issues, determine an optimal methodology for their resolution, and compile a data base for acceptable risk criteria development. The issues, spanning technical, psychological, and ethical dimensions, were categorized in seven major areas: (1) unplanned or accidental events, (2) present vs future risks, (3) institutional controls and retrievability, (4) dose-response mechanism and uncertainty, (5) spatial distribution of exposed populations, (6) different types of nuclear wastes, and (7) public perception. The optimum methodology for developing ARC was determined to be multi-attribute decision analysis encompassing numerous specific techniques for choosing, from among several alternatives, the optimal course of action when the alternatives are constrained to meet specified attributes. The data base developed during the study comprises existing regulations and guidelines, maximum permissible dose, natural geologic hazards, nonradioactive hazardous waste practices, bioethical perspectives, and data from an opinion survey

Determination of acceptable risk criteria for nuclear waste management

Energy Technology Data Exchange (ETDEWEB)

Cohen, J.J.

1977-10-21

The initial phase of the work performed during FY 1977 consisted of performing a ''scoping'' study to define issues, determine an optimal methodology for their resolution, and compile a data base for acceptable risk criteria development. The issues, spanning technical, psychological, and ethical dimensions, were categorized in seven major areas: (1) unplanned or accidental events, (2) present vs future risks, (3) institutional controls and retrievability, (4) dose-response mechanism and uncertainty, (5) spatial distribution of exposed populations, (6) different types of nuclear wastes, and (7) public perception. The optimum methodology for developing ARC was determined to be multi-attribute decision analysis encompassing numerous specific techniques for choosing, from among several alternatives, the optimal course of action when the alternatives are constrained to meet specified attributes. The data base developed during the study comprises existing regulations and guidelines, maximum permissible dose, natural geologic hazards, nonradioactive hazardous waste practices, bioethical perspectives, and data from an opinion survey.

Nuclear waste in public acceptance

International Nuclear Information System (INIS)

Vastchenko, Svetlana V.

2003-01-01

The existing problem on a faithful acceptance of nuclear information by population is connected, to a considerable extent, with a bad nuclear 'reputation' because of a great amount of misrepresented and false information from 'the greens'. In contrast to a bare style of professionals often neglecting an emotional perception, a loud voice of 'the greens' appeals both to the head, and to the heart of the audience. People pattern their behaviour weakly on problems of safe application of different irradiation sources in industry, conditions of life, medicine and everyday life. Radiation danger of some sources is often exaggerated (computers, nuclear technologies, radiation treatment) and the danger of the others is, on the contrary, underestimated (nuclear and roentgen methods of diagnostics and medical treatment). The majority of our citizens do not know which level of radiation is normal and safe, which ways radioactive substances intake into the organism of a human being and how to diminish the dose load on the organism by simple measures. Only specialists can be orientated themselves in a great number of radiation units. Low level of knowledge of the population and false conceptions are connected with the fact that they are mainly informed about nuclear technologies from mass media, where the voice of 'Greenpeace' is loudly sounded, but they often give misrepresented and false information doing it in the very emotional form. In contrast to them, scientists-professionals often ignore a sensitive part of apprehending of information and do not attach importance to it. As a rule, the style of specialists is of a serious academician character when they meet with the public. People preconception to nuclear waste and distrust to a positive information concerning nuclear technologies are explained, to a considerable extent, by a bivalent type of thinking when people operate by two opposite conceptions only, such as 'there is' or 'there is not' (there is or there is not

Nuclear waste in public acceptance

Energy Technology Data Exchange (ETDEWEB)

Vastchenko, Svetlana V. [Joint Institute for Power and Nuclear Research - Sosny / National Academy of Science, A.K.Krasin Str., 99, Minsk 220109 (Belarus)

2003-07-01

The existing problem on a faithful acceptance of nuclear information by population is connected, to a considerable extent, with a bad nuclear 'reputation' because of a great amount of misrepresented and false information from 'the greens'. In contrast to a bare style of professionals often neglecting an emotional perception, a loud voice of 'the greens' appeals both to the head, and to the heart of the audience. People pattern their behaviour weakly on problems of safe application of different irradiation sources in industry, conditions of life, medicine and everyday life. Radiation danger of some sources is often exaggerated (computers, nuclear technologies, radiation treatment) and the danger of the others is, on the contrary, underestimated (nuclear and roentgen methods of diagnostics and medical treatment). The majority of our citizens do not know which level of radiation is normal and safe, which ways radioactive substances intake into the organism of a human being and how to diminish the dose load on the organism by simple measures. Only specialists can be orientated themselves in a great number of radiation units. Low level of knowledge of the population and false conceptions are connected with the fact that they are mainly informed about nuclear technologies from mass media, where the voice of 'Greenpeace' is loudly sounded, but they often give misrepresented and false information doing it in the very emotional form. In contrast to them, scientists-professionals often ignore a sensitive part of apprehending of information and do not attach importance to it. As a rule, the style of specialists is of a serious academician character when they meet with the public. People preconception to nuclear waste and distrust to a positive information concerning nuclear technologies are explained, to a considerable extent, by a bivalent type of thinking when people operate by two opposite conceptions only, such as 'there is

Systematic handling of requirements and conditions (in compliance with waste acceptance requirements for a radioactive waste disposal facility)

International Nuclear Information System (INIS)

Keyser, Peter; Helander, Anita

2012-01-01

This Abstract and presentation will demonstrate the need for a structured requirement management and draw upon experiences and development from SKB requirements data base and methodology, in addition to international guidelines and software tools. The presentation will include a discussion on how requirement management can be applied for the decommissioning area. The key issue in the decommissioning of nuclear facilities is the progressive removal of hazards, by stepwise decontamination and dismantling activities that have to be carried out safely and within the boundaries of an approved safety case. For decommissioning there exists at least two safety cases, one for the pre-disposal activities and one for the disposal facility, and a need for a systematic handling of requirements and conditions to safely manage the radioactive waste in the long term. The decommissioning safety case is a collection of arguments and evidence to demonstrate the safety of a decommissioning project. It also includes analyzing and updating the decommissioning safety case in accordance with the waste acceptance criteria's and the expected output, i.e. waste packages. It is a continuous process to confirm that all requirements have been met. On the other hand there is the safety case for a radioactive waste disposal facility, which may include the following processes and requirements: i) Integrating relevant scientific (and other) information in a structured, traceable and transparent way and, thereby, developing and demonstrating an understanding of the potential behavior and performance of the disposal system; ii) Identifying uncertainties in the behavior and performance of the disposal system, describing the possible significance of the uncertainties, and identifying approaches for the management of significant uncertainties; iii) Demonstrating long-term safety and providing reasonable assurance that the disposal facility will perform in a manner that protects human health and the

TECHNO – ECONOMIC ACCEPTABILITY ANALISYS OF WASTE DISPOSAL BY INJECTION INTO APPROPRIATE FORMATION

Directory of Open Access Journals (Sweden)

Vladislav Brkić

2013-12-01

Full Text Available During exploration and production of oil and natural gas, various types of waste must be disposed in a permanent and safe way. There is a range of methods for processing and disposal of waste, such as disposal into landfills, solidification, namely chemical stabilization, thermal processing, appropriate formation injections uncovered by a deep well, disposal into salt domes and bioremediation. The method of waste disposal into appropriate formations is a method where strict geological and technical criteria must be satisfied when applied. A fundamental scientific hypothesis has been formulated whereby economic acceptability of the waste injection method, as a main method for waste disposal, is to be shown by an economic evaluation. The results of this research are relevant since there has been an intention in Croatia and worldwide to abandon wells permanently due to oil and gas reservoirs depletion and therefore it is essential to estimate economic impacts of the waste injection method application. In that way, profitability of using existing wells for waste disposal in oil industry has been increased, leading to the improvement of petroleum company’s business activities (the paper is published in Croatian.

Contested environmental policy infrastructure: Socio-political acceptance of renewable energy, water, and waste facilities

International Nuclear Information System (INIS)

Wolsink, Maarten

2010-01-01

The construction of new infrastructure is hotly contested. This paper presents a comparative study on three environmental policy domains in the Netherlands that all deal with legitimising building and locating infrastructure facilities. Such infrastructure is usually declared essential to environmental policy and claimed to serve sustainability goals. They are considered to serve (proclaimed) public interests, while the adverse impact or risk that mainly concerns environmental values as well is concentrated at a smaller scale, for example in local communities. The social acceptance of environmental policy infrastructure is institutionally determined. The institutional capacity for learning in infrastructure decision-making processes in the following three domains is compared: 1.The implementation of wind power as a renewable energy innovation; 2.The policy on space-water adaptation, with its claim to implement a new style of management replacing the current practice of focusing on control and 'hard' infrastructure; 3.Waste policy with a focus on sound waste management and disposal, claiming a preference for waste minimization (the 'waste management hierarchy'). All three cases show a large variety of social acceptance issues, where the appraisal of the impact of siting the facilities is confronted with the desirability of the policies. In dealing with environmental conflict, the environmental capacity of the Netherlands appears to be low. The policies are frequently hotly contested within the process of infrastructure decision-making. Decision-making on infrastructure is often framed as if consensus about the objectives of environmental policies exists. These claims are not justified, and therefore stimulating the emergence of environmental conflicts that discourage social acceptance of the policies. Authorities are frequently involved in planning infrastructure that conflicts with their officially proclaimed policy objectives. In these circumstances, they are

Preliminary criticality study supporting transuranic waste acceptance into the plasma hearth process

International Nuclear Information System (INIS)

Slate, L.J.; Santee, G.E. Jr.

1996-01-01

This study documents preliminary scoping calculations to address criticality issues associated with the processing of transuranic (TRU) waste and TRU mixed waste in the Plasma Hearth Process (PHP) Test Project. To assess the criticality potential associated with processing TRU waste, the process flow in the PHP is evaluated to identify the stages where criticality could occur. A criticality analysis methodology is then formulated to analyze the criticality potential. Based on these analyses, TRU acceptance criteria can be defined for the PHP. For the current level of analysis, the methodology only assesses the physical system as designed and does not address issues associated with the criticality double contingency principle. The analyses suggest that criticality within the PHP system and within the planned treatment residue (stag) containers does not pose a criticality hazard even when processing waste feed drums containing a quantity of TRU greater than would be reasonably expected. The analyses also indicate that the quantity of TRU that can be processed during each batch is controlled by moving and storage conditions for the resulting slag collection drums

Preliminary Assessment of the Hanford Tank Waste Feed Acceptance and Product Qualification Programs

Energy Technology Data Exchange (ETDEWEB)

Herman, C. C.; Adamson, Duane J.; Herman, D. T.; Peeler, David K.; Poirier, Micheal R.; Reboul, S. H.; Stone, M. E.; Peterson, Reid A.; Chun, Jaehun; Fort, James A.; Vienna, John D.; Wells, Beric E.

2013-04-01

The U.S. Department of Energy Office of Environmental Management (EM) is engaging the national laboratories to provide the scientific and technological rigor to support EM program and project planning, technology development and deployment, project execution, and assessment of program outcomes. As an early demonstration of this new responsibility, Savannah River National Laboratory (SRNL) and Pacific Northwest National Laboratory (PNNL) have been chartered to implement a science and technology program addressing Hanford Tank waste feed acceptance and product qualification. As a first step, the laboratories examined the technical risks and uncertainties associated with the planned waste feed acceptance and qualification testing for Hanford tank wastes. Science and technology gaps were identified for work associated with 1) feed criteria development with emphasis on identifying the feed properties and the process requirements, 2) the Tank Waste Treatment and Immobilization Plant (WTP) process qualification program, and 3) the WTP HLW glass product qualification program. Opportunities for streamlining the accetpance and qualification programs were also considered in the gap assessment. Technical approaches to address the science and technology gaps and/or implement the opportunities were identified. These approaches will be further refined and developed as strong integrated teams of researchers from national laboratories, contractors, industry, and academia are brought together to provide the best science and technology solutions. Pursuing the identified approaches will have immediate and long-term benefits to DOE in reducing risks and uncertainties associated with tank waste removal and preparation, transfers from the tank farm to the WTP, processing within the WTP Pretreatment Facility, and in producing qualified HLW glass products. Additionally, implementation of the identified opportunities provides the potential for long-term cost savings given the anticipated

Impact of community engagement on public acceptance towards waste-to-energy incineration projects: Empirical evidence from China.

Science.gov (United States)

Liu, Yong; Sun, Chenjunyan; Xia, Bo; Cui, Caiyun; Coffey, Vaughan

2018-02-20

As one of the most popular methods for the treatment of municipal solid waste (MSW), waste-to-energy (WTE) incineration offers effective solutions to deal with the MSW surge and globe energy issues. Nevertheless, the construction of WTE facilities faces considerable and strong opposition from local communities due to the perceived potential risks. The present study aims to understand whether, and how, community engagement improves local residents' public acceptance towards waste-to-energy (WTE) incineration facilities using a questionnaire survey conducted with nearby residents of two selected WTE incineration plants located in Zhejiang province, China. The results of data analysis using Structural Equation Modeling (SEM) reveal that firstly, a lower level of public acceptance exists among local residents of over the age of 35, of lower education levels, living within 3 km from the WTE Plant and from WTE incineration Plants which are under construction. Secondly, the public trust of local government and other authorities was positively associated with the public acceptance of the WTE incineration project, both directly and indirectly based on perceived risk. Thirdly, community engagement can effectively enhance public trust in local government and other authorities related to the WTE incineration project. The findings contribute to the literature on MSW treatment policy-making and potentially hazardous facility siting, by exploring the determinants of public acceptance towards WTE incineration projects. Copyright © 2018 Elsevier Ltd. All rights reserved.

Requirements on radioactive waste for disposal (waste acceptance requirements as of February 2017). Konrad repository; Anforderungen an endzulagernde radioaktive Abfaelle (Endlagerungsbedingungen, Stand: Februar 2017). Endlager Konrad

Energy Technology Data Exchange (ETDEWEB)

Kugel, Karin; Moeller, Kai (eds.)

2017-02-10

The Bundesamt fuer Strahlenschutz (BfS - Federal Office for Radiation Protection) has established waste acceptance requirements for the Konrad repository. These requirements were developed on the basis of the results of a site-specific safety assessment. They include general requirements on waste packages and specific requirements on waste forms and packagings as well as limitations for activities of individual radionuclides and limitations to masses of non-radioactive harmful substances. Requirements on documentation and delivery of waste packages were additionally included.

NUMERICAL FLOW AND TRANSPORT SIMULATIONS SUPPORTING THE SALTSTONE FACILITY PERFORMANCE ASSESSMENT

Energy Technology Data Exchange (ETDEWEB)

Flach, G.

2009-02-28

The Saltstone Disposal Facility Performance Assessment (PA) is being revised to incorporate requirements of Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA), and updated data and understanding of vault performance since the 1992 PA (Cook and Fowler 1992) and related Special Analyses. A hybrid approach was chosen for modeling contaminant transport from vaults and future disposal cells to exposure points. A higher resolution, largely deterministic, analysis is performed on a best-estimate Base Case scenario using the PORFLOW numerical analysis code. a few additional sensitivity cases are simulated to examine alternative scenarios and parameter settings. Stochastic analysis is performed on a simpler representation of the SDF system using the GoldSim code to estimate uncertainty and sensitivity about the Base Case. This report describes development of PORFLOW models supporting the SDF PA, and presents sample results to illustrate model behaviors and define impacts relative to key facility performance objectives. The SDF PA document, when issued, should be consulted for a comprehensive presentation of results.

Participatory approach, acceptability and transparency of waste management LCAs: case studies of Torino and Cuneo.

Science.gov (United States)

Blengini, Gian Andrea; Fantoni, Moris; Busto, Mirko; Genon, Giuseppe; Zanetti, Maria Chiara

2012-09-01

The paper summarises the main results obtained from two extensive applications of Life Cycle Assessment (LCA) to the integrated municipal solid waste management systems of Torino and Cuneo Districts in northern Italy. Scenarios with substantial differences in terms of amount of waste, percentage of separate collection and options for the disposal of residual waste are used to discuss the credibility and acceptability of the LCA results, which are adversely affected by the large influence of methodological assumptions and the local socio-economic constraints. The use of site-specific data on full scale waste treatment facilities and the adoption of a participatory approach for the definition of the most sensible LCA assumptions are used to assist local public administrators and stakeholders showing them that LCA can be operational to waste management at local scale. Copyright © 2012 Elsevier Ltd. All rights reserved.

Energy from waste: a wholly acceptable waste-management solution

International Nuclear Information System (INIS)

Porteous, A.

1997-01-01

This paper briefly reviews the 'waste management hierarchy' and why it should be treated as a checklist and not a piece of unquestioning dogma. The role of energy from waste (EfW) is examined in depth to show that it is a rigorous and environmentally sound waste-management option which complements other components of the waste-management hierarchy and assists resource conservation. (Copyright (c) 1997 Elsevier Science B.V., Amsterdam. All rights reserved.)

Assessment of national systems for obtaining local acceptance of waste management siting and routing activities

Energy Technology Data Exchange (ETDEWEB)

Paige, H.W.; Lipman, D.S.; Owens, J.E.

1980-07-01

There is a rich mixture of formal and informal approaches being used in our sister nuclear democracies in their attempts to deal with the difficulties of obtaining local acceptance for siting of waste management facilities and activities. Some of these are meeting with a degree of success not yet achieved in the US. Although this survey documents and assesses many of these approaches, time did not permit addressing in any detail their relevance to common problems in the US. It would appear the US could benefit from a periodic review of the successes and failures of these efforts, including analysis of their applicability to the US system. Of those countries (Germany, Sweden, Switzerland, Japan, Belgium, and the US) who are working to a time table for the preparation of a high-level waste (HLW) repository, Germany is the only country to have gained local siting acceptance for theirs. With this (the most difficult of siting problems) behind them they appear to be in the best overall condition relative to waste management progress and plans. This has been achieved without a particularly favorable political structure, made up for by determination on the part of the political leadership. Of the remaining three countries studied (France, UK and Canada) France, with its AVM production facility, is clearly the world leader in the HLW immobilization aspect of waste management. France, Belgium and the UK appear to have the least favorable political structures and environments for arriving at waste management decisions. US, Switzerland and Canada appear to have the least favorable political structures and environments for arriving at waste management decisions.

Hanford Waste Vitrification Plant: Preliminary description of waste form and canister

International Nuclear Information System (INIS)

Mitchell, D.E.

1986-01-01

In July 1985, the US Department of Energy's Office of Civilian Radioactive Waste Management established the Waste Acceptance Process as the means by which defense high-level waste producers, such as the Hanford Waste Vitrification Plant, will develop waste acceptance requirements with the candidate geologic repositories. A complete description of the Waste Acceptance Process is contained in the Preliminary Hanford Waste Vitrification Plant Waste Form Qualification Plan. The Waste Acceptance Process defines three documents that high-level waste producers must prepare as a part of the process of assuming that a high-level waste product will be acceptable for disposal in a geologic repository. These documents are the Description of Waste Form and Canister, Waste Compliance Plan, and Waste Qualification Report. This document is the Hanford Waste Vitrification Plant Preliminary Description of Waste Form and Canister for disposal of Neutralized Current Acid Waste. The Waste Acceptance Specifications for the Hanford Waste Vitrification Plant have not yet been developed, therefore, this document has been structured to corresponds to the Waste Acceptance Preliminary Specifications for the Defense Waste Processing Facility High-Level Waste Form. Not all of the information required by these specifications is appropriate for inclusion in this Preliminary Description of Waste Form and Canister. Rather, this description is limited to information that describes the physical and chemical characteristics of the expected high-level waste form. The content of the document covers three major areas: waste form characteristics, canister characteristics, and canistered waste form characteristics. This information will be used by the candidate geologic repository projects as the basis for preliminary repository design activities and waste form testing. Periodic revisions are expected as the Waste Acceptance Process progresses

Saltstone SDU6 Modeling Study

International Nuclear Information System (INIS)

Lee, Si Y.; Hyun, Sinjae

2013-01-01

A new disposal unit, designated as Saltstone Disposal Unit 6 (SDU6), is being designed for support of site accelerated closure goals and salt waste projections identified in the new Liquid Waste System Plan. The unit is a cylindrical disposal cell of 375 ft in diameter and 43 ft in height, and it has a minimum 30 million gallons of capacity. SRNL was requested to evaluate the impact of an increased grout placement height on the flow patterns radially spread on the floor and to determine whether grout quality is impacted by the height. The primary goals of the work are to develop the baseline Computational Fluid Dynamics (CFD) model and to perform the evaluations for the flow patterns of grout material in SDU6 as a function of elevation of grout discharge port and grout rheology. Two transient grout models have been developed by taking a three-dimensional multiphase CFD approach to estimate the domain size of the grout materials radially spread on the facility floor and to perform the sensitivity analysis with respect to the baseline design and operating conditions such as elevation height of the discharge port and fresh grout properties. For the CFD modeling calculations, air-grout Volume of Fluid (VOF) method combined with Bingham plastic and time-dependent grout models were used for examining the impact of fluid spread performance for the initial baseline configurations and to evaluate the impact of grout pouring height on grout quality. The grout quality was estimated in terms of the air volume fraction for the grout layer formed on the SDU6 floor, resulting in the change of grout density. The study results should be considered as preliminary scoping analyses since benchmarking analysis is not included in this task scope. Transient analyses with the Bingham plastic model were performed with the FLUENTTM code on the high performance parallel computing platform in SRNL. The analysis coupled with a transient grout aging model was performed by using ANSYS-CFX code

Present state of the design and realization of regional radioactive waste depositories and waste acceptance criteria for disposal

International Nuclear Information System (INIS)

Kortus, J.

1988-01-01

Surface type regional depositories for radioactive wastes from nuclear power plants are described in detail. The depository of the Mochovce nuclear power plant is located near the plant, that of the Dukovany nuclear power plant is directly on the premises of the plant. Particular attention is paid to the design of the monolithic reinforced concrete pits, draining of rainwater from their surface, draining of seeping rainwater from the pit environment by means of a double drainage system, and insulation of the pits against water. The construction of the Mochovce depository started in 1987; some experience gained from this activity is presented. The radioactive waste acceptance criteria for depositories of this kind, based on safety analysis, are given. (author). 2 figs

Perceived risks of nuclear fuel waste disposal: trust, compensation, and public acceptance in Canada

International Nuclear Information System (INIS)

Hine, D.W.; Summers, C.

1996-01-01

AECL's recommendation to place the high-level radioactive waste in corrosion resistant containers and bury it in underground vaults several hundred metres deep in the rock of the Canadian shield is presently under federal review. If and when the disposal concept is approved by the federal review panel, a search will begin for a suitable host community. Given that siting guidelines prevent the government from unilaterally imposing the waste on a reluctant community, identifying a suitable site may represent the single greatest obstacle to successfully implementing the disposal concept. Even if the concept is approved by the review panel, it may be very difficult to find a community that is willing to accept the waste. In the US, efforts to site an underground disposal facility for high-level nuclear waste at Yucca Mountain has run into strong opposition from local residents and politicians, resulting in long delays and major cost overruns

Framing ethical acceptability: a problem with nuclear waste in Canada.

Science.gov (United States)

Wilding, Ethan T

2012-06-01

Ethical frameworks are often used in professional fields as a means of providing explicit ethical guidance for individuals and institutions when confronted with ethically important decisions. The notion of an ethical framework has received little critical attention, however, and the concept subsequently lends itself easily to misuse and ambiguous application. This is the case with the 'ethical framework' offered by Canada's Nuclear Waste Management Organization (NWMO), the crown-corporation which owns and is responsible for the long-term management of Canada's high-level nuclear fuel waste. It makes a very specific claim, namely that it is managing Canada's long-lived radioactive nuclear fuel waste in an ethically responsible manner. According to this organization, what it means to behave in an ethically responsible manner is to act and develop policy in accordance with its ethical framework. What, then, is its ethical framework, and can it be satisfied? In this paper I will show that the NWMO's ethical and social framework is deeply flawed in two respects: (a) it fails to meet the minimum requirements of a code of ethic or ethical framework by offering only questions, and no principles or rules of conduct; and (b) if posed as principles or rules of conduct, some of its questions are unsatisfiable. In particular, I will show that one of its claims, namely that it seek informed consent from individuals exposed to risk of harm from nuclear waste, cannot be satisfied as formulated. The result is that the NWMO's ethical framework is not, at present, ethically acceptable.

Cross-Site Transfer System at Hanford: long-term strategy for waste acceptance

International Nuclear Information System (INIS)

Shekarriz, A.; Onishi, Y.; Smith, P.A.; Sterner, M.; Rector, D.R.; Virden, J.

1997-02-01

This report summarizes results of a technical panel review of the current methodology for accepting waste for transport through the Hanford Replacement Cross-Site Transfer System (RCSTS), which was constructed to replace the existing pipelines that hydraulically connect the 200 West and 200 East areas. This report is a complement to an existing document (Hudson 1996); the methodology proposed in that document was refined based on panel recommendations. The refinements were focused around predicting and preventing the 3 main modes suspected of plugging the existing CSTS: precipitation, gelation, particle dropout/settling. The proposed analysis will require integration of computer modeling and laboratory experiments to build a defensible case for transportability of a proposed slurry composition for a given tank. This will be validated by recirculating actual tank waste, in-tank and in-farm, prior to transport. The panel's recommendation was that the probability of success of waste transfer would be greatly improved by integrating the predictive analysis with real-time control during RCSTS operation. The methodology will be optimized

Development of comprehensive waste acceptance criteria for commercial nuclear waste

International Nuclear Information System (INIS)

O'Hara, F.A.; Miller, N.E.; Ausmus, B.S.; Yates, K.R.; Means, J.L.; Christensen, R.N.; Kulacki, F.A.

1979-01-01

A detailed methodology is presented for the identification of the characteristics of commercial nuclear waste which may require criteria. This methodology is analyzed as a six-step process which begins with identification of waste operations and proceeds until the waste characteristics affecting the potential release of radionuclides are determined. All waste types and operations were analyzed using the methodology presented. Several illustrative example are included. It is found that thirty-three characteristics can be identified as possibly requiring criteria

Results For The Third Quarter 2010 Tank 50 WAC Slurry Sample: Chemical And Radionuclide Contaminant Results

International Nuclear Information System (INIS)

Reigel, M.; Bibler, N.

2010-01-01

This report details the chemical and radionuclide contaminant results for the characterization of the 2010 Third Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC). Information from this characterization will be used by Liquid Waste Operations (LWO) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System. The following conclusions are drawn from the analytical results provided in this report: (i) The concentrations of the reported chemical and radioactive contaminants were less than their respective WAC targets or limits unless noted in this section. (ii) The reported detection limits for 94 Nb, 247 Cm and 249 Cf are above the requested limits from Reference 4. However, they are below the limits established in Reference 3. (iii) The reported detection limit for 242m Am is greater than the requested limit from Attachment 8.4 of the WAC. (iv) The reported detection limit for Isopar L is greater than the limit from Table 3 of the WAC. (v) The reported concentration of Isopropanol is greater than the limit from Table 4 of the WAC. (vi) Isopar L and Norpar 13 have limited solubility in aqueous solutions making it difficult to obtain consistent and reliable sub-samples. The values reported in this memo are the concentrations in the sub-sample as detected by the GC/MS; however, the results may not accurately represent the concentrations of the analytes in Tank 50.

Secondary Waste Cast Stone Waste Form Qualification Testing Plan

Energy Technology Data Exchange (ETDEWEB)

Westsik, Joseph H.; Serne, R. Jeffrey

2012-09-26

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). Cast Stone – a cementitious waste form, has been selected for solidification of this secondary waste stream after treatment in the ETF. The secondary-waste Cast Stone waste form must be acceptable for disposal in the IDF. This secondary waste Cast Stone waste form qualification testing plan outlines the testing of the waste form and immobilization process to demonstrate that the Cast Stone waste form can comply with the disposal requirements. Specifications for the secondary-waste Cast Stone waste form have not been established. For this testing plan, Cast Stone specifications are derived from specifications for the immobilized LAW glass in the WTP contract, the waste acceptance criteria for the IDF, and the waste acceptance criteria in the IDF Permit issued by the State of Washington. This testing plan outlines the testing needed to demonstrate that the waste form can comply with these waste form specifications and acceptance criteria. The testing program must also demonstrate that the immobilization process can be controlled to consistently provide an acceptable waste form product. This testing plan also outlines the testing needed to provide the technical basis for understanding the long-term performance of the waste form in the disposal environment. These waste form performance data are needed to support performance assessment analyses of the long-term environmental impact of the secondary-waste Cast Stone waste form in the IDF

Public Acceptance of Low-Level Waste Disposal Critical to the Nuclear Renaissance

International Nuclear Information System (INIS)

Sonny Goldston, W.T.

2009-01-01

The disposal of various Low-Level Waste (LLW) forms projected to result from the operation of a pilot or large scale Advanced Fuel Cycle Initiative Programs' (formally known as Global Nuclear Energy Partnership (GNEP)) reprocessing and vitrification plants requires the DOE LLW program and regulatory structure to be utilized in its present form due to the limited availability of Nuclear Regulatory Commission licensed commercial LLW disposal facilities to handle wastes with radionuclide concentrations that are greater than Nuclear Regulatory Commission (NRC) Class A limits. This paper will describe the LLW forms and the regulatory structures and facilities available to dispose of this waste. Then the paper discusses the necessity of an excellent public involvement program to ensure the success of an effective technical solution. All of the decisions associated with the management of these wastes are of interest to the public and successful program implementation would be impossible without including the public up-front in the program formulation. Serious problems can result if program decisions are made without public involvement, and if the public is informed after key decisions are made. This paper will describe the regulatory and public involvement program and their effects on the decisions concerning the disposal of Low-Level Radioactive Waste (LLW) at the Savannah River Site (SRS). An extensive public communications effort resulted in endorsement of changes in disposal practices by the SRS Citizens Advisory Board that was critical to the success of the program. A recommendation will be made to install a public involvement program that is similar to the SRS Citizens Advisory Board in order to ensure the success of the AFCI programs in view of the limited availability to handle the wastes from the program and the public acceptance of change that will be required. (authors)

A data base and a standard material for use in acceptance testing of low-activity waste products

International Nuclear Information System (INIS)

Wolf, S.F.; Ebert, W.L.; Luo, J.S.; Strachan, D.M.

1998-04-01

The authors have conducted replicate dissolution tests following the product consistency test (PCT) procedure to measure the mean and standard deviation of the solution concentrations of B, Na, and Si at various combinations of temperature, duration, and glass/water mass ratio. Tests were conducted with a glass formulated to be compositionally similar to low-activity waste products anticipated for Hanford to evaluate the adequacy of test methods that have been designated in privatization contracts for use in product acceptance. An important finding from this set of tests is that the solution concentrations generated in tests at 20 C will likely be too low to measure the dissolution rates of waste products reliably. Based on these results, the authors recommend that the acceptance test be conducted at 40 C. Tests at 40 C generated higher solution concentrations, were more easily conducted, and the measured rates were easily related to those at 20 C. Replicate measurements of other glass properties were made to evaluate the possible use of LRM-1 as a standard material. These include its composition, homogeneity, density, compressive strength, the Na leachability index with the ANSI/ANS 16.1 leach test, and if the glass is characteristically hazardous with the toxicity characteristic leach procedure. The values of these properties were within the acceptable limits identified for Hanford low-activity waste products. The reproducibility of replicate tests and analyses indicates that the glass would be a suitable standard material

A revised oceanographic model to calculate the limiting capacity of the ocean to accept radioactive waste

International Nuclear Information System (INIS)

Webb, G.A.M.; Grimwood, P.D.

1976-12-01

This report describes an oceanographic model which has been developed for the use in calculating the capacity of the oceans to accept radioactive wastes. One component is a relatively short-term diffusion model which is based on that described in an earlier report (Webb et al., NRPB-R14(1973)), but which has been generalised to some extent. Another component is a compartment model which is used to calculate long-term widespread water concentrations. This addition overcomes some of the short comings of the earlier diffusion model. Incorporation of radioactivity into deep ocean sediments is included in this long-term model as a removal mechanism. The combined model is used to provide a conservative (safe) estimate of the maximum concentrations of radioactivity in water as a function of time after the start of a continuous disposal operation. These results can then be used to assess the limiting capacity of an ocean to accept radioactive waste. (author)

Gamma-ray spectrometry combined with acceptable knowledge (GSAK). A technique for characterization of certain remote-handled transuranic (RH-TRU) wastes. Part 1. Methodology and techniques

International Nuclear Information System (INIS)

Hartwell, J.K.; McIlwain, M.E.

2005-01-01

Gamma-ray spectrometry combined with acceptable knowledge (GSAK) is a technique for the characterization of certain remote-handled transuranic (RH-TRU) wastes. GSAK uses gamma-ray spectrometry to quantify a portion of the fission product inventory of RH-TRU wastes. These fission product results are then coupled with calculated inventories derived from acceptable process knowledge to characterize the radionuclide content of the assayed wastes. GSAK has been evaluated and tested through several test exercises. GSAK approach is described, while test results are presented in Part II. (author)

The DWPF strategy for producing an acceptable product

International Nuclear Information System (INIS)

Goldston, W.T.; Plodinec, M.J.

1991-01-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will convert the 130 million liters of high-level nuclear waste at SRS into stable borosilicate glass. Production of canistered waste forms by the DWPF is scheduled to begin well before submission of the license application for the first repository. The Department of Energy has defined waste acceptance specifications to ensure that DWPF canistered waste forms will be acceptable for eventual disposal. To ensure that canistered waste forms meet those specifications, a program is being carried out to qualify the waste form and those aspects of the production process which affect product quality. This program includes: Pre-production qualification testing of simulated and actual waste forms; Disciplined demonstrations of the ability to produce an acceptable product during startup testing; and Application of a rigorous product control program during production

Final waste classification and waste form technical position papers

International Nuclear Information System (INIS)

1983-05-01

The waste classification technical position paper describes overall procedures acceptable to NRC staff which may be used by licensees to determine the presence and concentrations of the radionuclides listed in section 61.55, and thereby classifying waste for near-surface disposal. This technical position paper also provides guidance on the types of information which should be included in shipment manifests accompanying waste shipments to near-surface disposal facilities. The technical position paper on waste form provides guidance to waste generators on test methods and results acceptable to NRC staff for implementing the 10 CFR Part 61 waste form requirements. It can be used as an acceptable approach for demonstrating compliance with the 10 CFR Part 61 waste structural stability criteria. This technical position paper includes guidance on processing waste into an acceptable stable form, designing acceptable high-integrity containers, packaging cartridge filters, and minimizing radiation effects on organic ion-exchange resins. The guidance in the waste form technical position paper may be used by licensees as the basis for qualifying process control programs to meet the waste form stability requirements, including tests which can be used to demonstrate resistance to degradation arising from the effects of compression, moisture, microbial activity, radiation, and chemical changes. Generic test data (e.g., topical reports prepared by vendors who market solidification technology) may be used for process control program qualification where such generic data is applicable to the particular types of waste generated by a licensee

Radioactive waste and public acceptance

International Nuclear Information System (INIS)

Perkins, B.

1977-01-01

Radioactive waste just happens to be the major issue in the public eye now--it could be replaced by another issue later. A survey is quoted to prove that wastes are not really one of the burning national issues of the day. The people opposing the nuclear program cannot be said to represent the public. The taste of the press for the melodramatic is pointed out. The issue needs to be presented with the proper perspective, in the context of the benefits and risks of nuclear power

Tests of a system to exclude roots from buried radioactive waste in a warm, humid climate

International Nuclear Information System (INIS)

Murphy, C.E. Jr.; Corey, J.C.; Adriano, D.C.; Decker, O.D.; Griggs, R.D.

1989-01-01

Vegetation is commonly used to stabilize the ground covering buried waste sites. However, constituents of buried waste can be brought to the surface if the waste is penetrated by plant roots. An ideal waste burial system would allow the use of vegetation to stabilize the soil above the buried waste but would exclude roots from the waste. One system that shows considerable promise is a slow release encapsulation of a root growth inhibitor (Trifluralin). Projected lifetimes of the capsule are in the order of 100 years. The capsule is bonded to a geotextile, which provides an easy means of distributing the capsule evenly over the area to be protected. Vegetation grown in the soil above the barrier has provided good ground cover, although some decrease in growth has been found in some species. Of the species tested the sensitivity to the biobarrier, as measured by the distance root growth stops near the barrier, is bamboo> bahia grass> bermuda grass> soybean. Potential uses for the biobarrier at the Savannah River Site (SRS) include the protection of clay caps over buried, low-level saltstone and protection of gravel drains and clay caps over decommissioned seepage basins. Trails of the biobarrier as part of waste site caps are scheduled to begin during the next 12 months

Gamma-ray spectrometry combined with acceptable knowledge (GSAK). A technique for characterization of certain remote-handled transuranic (RH-TRU) wastes. Part 2. Testing and results

International Nuclear Information System (INIS)

Hartwell, J.K.; McIlwain, M.E.

2005-01-01

Gamma-ray spectrometry combined with acceptable knowledge (GSAK) is a technique for the characterization of certain remote-handled transuranic (RH-TRU) wastes. GSAK uses gamma-ray spectrometry to quantify a portion of the fission product inventory of RH-TRU wastes. These fission product results are then coupled with calculated inventories derived from acceptable process knowledge to characterize the radionuclide content of the assayed wastes. GSAK has been evaluated and tested through several test exercises. These tests and their results are described; while the former paper in this issue presents the methodology, equipment and techniques. (author)

Statistical experimental design for saltstone mixtures

International Nuclear Information System (INIS)

Harris, S.P.; Postles, R.L.

1991-01-01

We used a mixture experimental design for determining a window of operability for a process at the Savannah River Site Defense Waste Processing Facility (DWPF). The high-level radioactive waste at the Savannah River Site is stored in large underground carbon steel tanks. The waste consists of a supernate layer and a sludge layer. 137 Cs will be removed from the supernate by precipitation and filtration. After further processing, the supernate layer will be fixed as a grout for disposal in concrete vaults. The remaining precipitate will be processed at the DWPF with treated waste tank sludge and glass-making chemicals into borosilicate glass. The leach rate properties of the supernate grout, formed from various mixes of solidified salt waste, needed to be determined. The effective diffusion coefficients for NO 3 and Cr were used as a measure of leach rate. Various mixes of cement, Ca(OH) 2 , salt, slag and flyash were used. These constituents comprise the whole mix. Thus, a mixture experimental design was used

1991 Acceptance priority ranking

International Nuclear Information System (INIS)

1991-12-01

The Standard Contract for Disposal of Spent Nuclear Fuel and/or High- Level Radioactive Waste (10 CFR Part 961) that the Department of Energy (DOE) has executed with the owners and generators of civilian spent nuclear fuel requires annual publication of the Acceptance Priority Ranking (APR). The 1991 APR details the order in which DOE will allocate Federal waste acceptance capacity. As required by the Standard Contract, the ranking is based on the age of permanently discharged spent nuclear fuel (SNF), with the owners of the oldest SNF, on an industry-wide basis, given the highest priority. the 1991 APR will be the basis for the annual allocation of waste acceptance capacity to the Purchasers in the 1991 Annual Capacity Report (ACR), to be issued later this year. This document is based on SNF discharges as of December 31, 1990, and reflects Purchaser comments and corrections, as appropriate, to the draft APR issued on May 15, 1991

Acceptance criteria considerations for miscellaneous wastes

International Nuclear Information System (INIS)

Irvine, A.R.; Forsberg, C.W.

1987-01-01

EPA standards set forth limitations regarding releases to the accessible environment adjacent to a geologic repository. The NRC criteria pertaining to waste form and engineered barrier performance place certain restrictions on the physical and chemical nature of the waste form and require substantially complete confinement of radioactivity until the high-heat-production period is past. After this period, the annual release of radionuclides from the waste package is normally limited to 1 part in 100,000 of the amounts calculated to be present at 1000-y decay. The regulation permits deviation from these criteria in exceptional circumstances. One such circumstance might be the absence of a significant perturbation in temperature around the stored waste. The lack of significant heat release will eliminate the hydrologic driving force for dispersal of radionuclides. Exceptional circumstances which potentially could justify a less stringent long-term release criterion are: small quantity of radioactivity, the nature of the radioactive species, and the nature of the geology in which the waste is to be emplaced. Because the MW after a suitable decay period have low heat release rates per unit volume, they apparently could be so emplaced in a repository that there would be no compelling need, according to the reasoning presented in 10 CFR 60, for a 1000-y container. Regarding attainment of the specified long-term release rate criterion, neither the solubility limits for the various waste forms nor the conductance of potential migration barriers are currently adequately characterized. The relatively small total heat generation rate for the MW in combination with the usual low volumetric heat generation rate apparently will allow application of migration barriers in a low temperature environment where barrier performance would be expected to be unchanged with time

High-level waste management technology program plan

Energy Technology Data Exchange (ETDEWEB)

Harmon, H.D.

1995-01-01

The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs.

High-level waste management technology program plan

International Nuclear Information System (INIS)

Harmon, H.D.

1995-01-01

The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs

Technical considerations associated with spent fuel acceptance. Final report

International Nuclear Information System (INIS)

Supko, E.M.

1996-06-01

This study was initiated by the Electric Power Research Institute (EPRI) to identify technical considerations associated with spent fuel acceptance and implementation of a waste management system that includes the use of transportable storage systems, and to serve as an opening dialogue among Standard Contract Holders and the department of Energy's Office of Civilian Radioactive Waste management (OCRWM) prior to the development of waste acceptance criteria or issuance of a Notice of Proposed Rulemaking by OCRWM to amend the Standard Contract. The original purpose of the Notice of Proposed Rulemaking was to address changes to the Standard Contract to implement a multi-purpose canister based system and to address other issues that were not adequately addressed in the standard contract. Even if DOE does not develop a multi-purpose canister based system for waste acceptance, it will still be necessary to develop waste acceptance criteria in order to accept spent fuel in transportable storage systems that are being deployed for at-reactor storage. In this study, technical issues associated with spent fuel acceptance will be defined and potential options and alternatives for resolution of technical considerations will be explored

Assessment of national systems for obtaining local siting acceptance of nuclear waste management facilities (October 1, 1985). Volume I. Political structure and formal system for obtaining approvals for siting waste management facilities

International Nuclear Information System (INIS)

Paige, H.W.; Numark, N.J.

1985-01-01

This report is the fourth in a series of periodic surveys of approaches and progress in other countries in dealing with the problems of obtaining local acceptance for siting of waste management facilities. This volume contains the following sections: Nation's political/industrial structure for obtaining waste management siting decisions; and Nation's formal legal procedure for obtaining necessary approvals for siting nuclear waste management facilities. Two of the countries visited, Finland and Sweden, have had major changes in the past two years in their formal/legal procedures for obtaining waste management siting decisions. (LM)

Derivation of quantitative acceptance criteria for disposal of radioactive waste to near surface facilities: Development and implementation of an approach for the post-closure phase

International Nuclear Information System (INIS)

Torres, C.

2000-01-01

The International Atomic Energy Agency has established a project to develop and illustrate, through practical examples, an approach that allows the derivation of quantitative waste acceptance criteria for near surface disposal of radioactive waste. The first phase focussed on the derivation of example post-closure safety waste acceptance criteria through the use of a safety assessment approach that allows for the derivation of values in a clear and well documented manner. The approach consists of five steps: the specification of the assessment context; the description of the disposal system; the development and justification of scenarios; the formulation and implementation of models; and the calculation and derivation of example values. The approach has been successfully used to derive activity values for the disposal of radioactive waste to illustrative near surface facilities. (author)

Central Waste Complex (CWC) Waste Analysis Plan

International Nuclear Information System (INIS)

ELLEFSON, M.D.

2000-01-01

The purpose of this waste analysis plan (WAP) is to document the waste acceptance process, sampling methodologies, analytical techniques, and overall processes that are undertaken for waste accepted for storage at the Central Waste Complex (CWC), which is located in the 200 West Area of the Hanford Facility, Richland, Washington. Because dangerous waste does not include the source special nuclear and by-product material components of mixed waste, radionuclides are not within the scope of this document. The information on radionuclides is provided only for general knowledge. This document has been revised to meet the interim status waste analysis plan requirements of Washington Administrative Code (WAC) 173 303-300(5). When the final status permit is issued, permit conditions will be incorporated and this document will be revised accordingly

Proceedings of the workshop on radioactive, hazardous, and/or mixed waste sludge management

International Nuclear Information System (INIS)

Lomenick, T.F.

1992-01-01

A workshop sponsored by the US Department of Energy (DOE) Field Office, Oak Ridge, was held on December 4--6, 1990, in Knoxville, Tennessee. The primary objective of the workshop was the exchange of information, experiences, solutions, and future plans of DOE and its prime contractors who are engaged in work on the packaging, grouting, storage, and transport of waste sludges. In addition, the group met with industrial participants in an open forum to discuss problems and needs in the management of these wastes and to learn of possible industrial experiences, approaches, and solutions, including demonstrations of potential tools and techniques. Topics discussed include the following: mixed waste sludge issue at the K-25 site; processing saltstone from waste streams at the Savannah River Plant; the Hanford Grout Treatment Facility; treatment of pond sludge at the Rocky Flats Plant; cement solidification of low-level radioactive sludge at the West Valley Demonstration Project; studies on the solidification of low-level radioactive wastes in cement at INEL; cement solidification systems at Los Alamos National Laboratory; emergency avoidance solidification campaign at ORNL; diffusion plant sludge storage problems at the Portsmouth Gaseous Diffusion Plant; the proposed fixation of sludge in cement at the feed materials production center; regulatory aspects of sludge management; and delisting efforts for K-1407-C pond sludges. Individual projects are processed separately for the data bases

Hanford Waste Vitrification Plant Project Waste Form Qualification Program Plan

International Nuclear Information System (INIS)

Randklev, E.H.

1993-06-01

The US Department of Energy has created a waste acceptance process to help guide the overall program for the disposal of high-level nuclear waste in a federal repository. This Waste Form Qualification Program Plan describes the hierarchy of strategies used by the Hanford Waste Vitrification Plant Project to satisfy the waste form qualification obligations of that waste acceptance process. A description of the functional relationship of the participants contributing to completing this objective is provided. The major activities, products, providers, and associated scheduling for implementing the strategies also are presented

Structural acceptance criteria for the evaulation of existing double-shell waste storage tanks located at the Hanford site, Richland, Washington

International Nuclear Information System (INIS)

Julyk, L.J.; Day, A.D.; Dyrness, A.D.; Moore, C.J.; Peterson, W.S.; Scott, M.A.; Shrivastava, H.P.; Sholman, J.S.; Watts, T.N.

1995-09-01

The structural acceptance criteria contained herein for the evaluation of existing underground double-shell waste storage tanks located at the Hanford Site is part of the Life Management/Aging Management Program of the Tank Waste Remediation System. The purpose of the overall life management program is to ensure that confinement of the waste is maintained over the required service life of the tanks. Characterization of the present condition of the tanks, understanding and characterization of potential degradation mechanisms, and development of tank structural acceptance criteria based on previous service and projected use are prerequisites to assessing tank integrity, to projecting the length of tank service, and to developing and applying prudent fixes or repairs. The criteria provided herein summarize the requirements for the analysis and structural qualification of the existing double-shell tanks for continued operation. Code reconciliation issues and material degradation under aging conditions are addressed. Although the criteria were developed for double-shell tanks, many of the provisions are equally applicable to single-shell tanks. However, the criteria do not apply to the evaluation of tank appurtenances and buried piping

Disposal of low-level and mixed low-level radioactive waste during 1990

International Nuclear Information System (INIS)

1993-08-01

Isotopic inventories and other data are presented for low-level radioactive waste (LLW) and mixed LLW disposed (and occasionally stored) during calendar year 1990 at commercial disposal facilities and Department of Energy (DOE) sites. Detailed isotopic information is presented for the three commercial disposal facilities located near Barnwell, SC, Richland, WA, and Beatty, NV. Less information is presented for the Envirocare disposal facility located near Clive, UT, and for LLW stored during 1990 at the West Valley site. DOE disposal information is included for the Savannah River Site (including the saltstone facility), Nevada Test Site, Los Alamos National Laboratory, Idaho National Engineering Laboratory, Hanford Site, Y-12 Site, and Oak Ridge National Laboratory. Summary information is presented about stored DOE LLW. Suggestions are made about improving LLW disposal data

Verifying generator waste certification: NTS waste characterization QA requirements

International Nuclear Information System (INIS)

Williams, R.E.; Brich, R.F.

1988-01-01

Waste management activities managed by the US Department of Energy (DOE) at the Nevada Test Site (NTS) include the disposal of low-level wastes (LLW) and mixed waste (MW), waste which is both radioactive and hazardous. A majority of the packaged LLW is received from offsite DOE generators. Interim status for receipt of MW at the NTS Area 5 Radioactive Waste Management Site (RWMS) was received from the state of Nevada in 1987. The RWMS Mixed Waste Management Facility (MWMF) is expected to be operational in 1988 for approved DOE MW generators. The Nevada Test Site Defense Waste Acceptance Criteria and Certification Requirements (NVO-185, Revision 5) delineates waste acceptance criteria for waste disposal at the NTS. Regulation of the hazardous component of mixed waste requires the implementation of US Environmental Protection Agency (EPA) requirements pursuant to the Resource Conservation and Recovery Act (RCRA). Waste generators must implement a waste certification program to provide assurance that the disposal site waste acceptance criteria are met. The DOE/Nevada Operations Office (NV) developed guidance for generator waste certification program plans. Periodic technical audits are conducted by DOE/NV to assess performance of the waste certification programs. The audit scope is patterned from the waste certification program plan guidance as it integrates and provides a common format for the applicable criteria. The criteria focus on items and activities critical to processing, characterizing, packaging, certifying, and shipping waste

Identification and understanding the factors affecting the public and political acceptance of long term storage of spent fuel and high-level radioactive wastes

International Nuclear Information System (INIS)

Gorea, Valica

2006-01-01

In the end of 2004, according to the information available to the IAEA, there were 440 nuclear reactors operating worldwide, with a total net capacity of 366.3 GW(e), 6 of them being connected to the grid in 2004 ( 2 in Ukraine, one each in China, Japan and the Russian Federation and a reconnection in Canada) by comparison with 2 connections and 2 re-connections in 2003. Also, in the end of 2004, 26 nuclear power plants were under construction with a total net capacity of 20.8 GW(e). The conclusion accepted by common consent is that the nuclear power is still in progress and represents one of the options for power security on long and middle term. If we refer to the nuclear fusion which will produce commercial electric power, over 30 - 40 years, in practically unlimited quantities, the above underlining becomes even more evident. Fortunately, besides the beneficent characteristics, such as: clean, stable as engendering and price, has also a negative characteristic, which generally breathes fear into the people: radioactive waste. A classification of the radioactive waste is not the target of this presentation. I just want to point that a nuclear power plant produces during the time spent fuel - long life high radioactive, generating heat. Another high radioactive waste have similar characteristics (HLW = High Level Waste) for which reason these two categories of wastes are treated together. The spent fuel and the High Level Waste are interim stored for cooling, for around 50 years, afterwards it is transferred to the final repository where it will be kept for hundreds of years, in the case of an open fuel cycle and this is also the case of Cernavoda NPP. Taking into consideration that the Cernavoda Unit 1 reaches the age of 10 years of commercial running during December 2006, it results that the issue of the final disposal is not such urgent as it looks. The objectives of long term management of radioactive waste are public health and protection of the environment

Approaches to gaining public acceptance of repository siting

International Nuclear Information System (INIS)

Numark, N.J.; Wonder, E.F.

1989-01-01

An eight-country survey reveals a diversity of strategies that have been followed for siting radioactive waste repositories, as well as a range of levels of public acceptance of siting efforts. Although the strategies are not necessarily interchangeable from country to country, certain inferences may be drawn from worldwide siting experience regarding ways to maximize public acceptance. Furthermore, waste management organizations in these countries have placed varying amounts of stock in technical review by outside experts and in a range of communications strategies as means of improving public acceptance. Our survey of worldwide experience also allows some general observations to be made regarding the effectiveness of these efforts. Combining a strategy that elevates public acceptance to part of the overall mission of siting a waste management facility with strategies for appropriate communications and external technical review may be necessary and sufficient for gaining improvements in public acceptance of proposed repository sites

Improving societal acceptance of rad waste management policy decisions: an approach based on complex intelligence

International Nuclear Information System (INIS)

Rao, Suman

2008-01-01

In today's context elaborate public participation exercises are conducted around the world to elicit and incorporate societal risk perceptions into nuclear policy Decision-Making. However, on many occasions, such as in the case of rad waste management, the society remains unconvinced about these decisions. This naturally leads to the questions: are techniques for incorporating societal risk perceptions into the rad waste policy decision making processes sufficiently mature? How could societal risk perceptions and legal normative principles be better integrated in order to render the decisions more equitable and convincing to society? Based on guidance from socio-psychological research this paper postulates that a critical factor for gaining/improving societal acceptance is the quality and adequacy of criteria for option evaluation that are used in the policy decision making. After surveying three rad waste public participation cases, the paper identifies key lacunae in criteria abstraction processes as currently practiced. A new policy decision support model CIRDA: Complex Intelligent Risk Discourse Abstraction model that is based on the heuristic of Risk-Risk Analysis is proposed to overcome these lacunae. CIRDA's functionality of rad waste policy decision making is modelled as a policy decision-making Abstract Intelligent Agent and the agent program/abstraction mappings are presented. CIRDA is then applied to a live (U.K.) rad waste management case and the advantages of this method as compared to the Value Tree Method as practiced in the GB case are demonstrated. (author)

Acceptance, Tolerance, Participation

International Nuclear Information System (INIS)

1993-01-01

The problem of radioactive waste management from an ethical and societal viewpoint was treated in this seminar, which had participants from universities (social, theological, philosophical and science institutes), waste management industry, and regulatory and controlling authorities. After initial reviews on repository technology, policies and schedules, knowledge gaps, and ethical aspects on decision making under uncertainty, four subjects were treated in lectures and discussions: Democratic collective responsibility, Handling threats in democratic decision making, Waste management - a technological operation with a social dimension, Acceptance and legitimity. Lectures with comments and discussions are collected in this report

Waste Determination Equivalency - 12172

Energy Technology Data Exchange (ETDEWEB)

Freeman, Rebecca D. [Savannah River Remediation (United States)

2012-07-01

The Savannah River Site (SRS) is a Department of Energy (DOE) facility encompassing approximately 800 square kilometers near Aiken, South Carolina which began operations in the 1950's with the mission to produce nuclear materials. The SRS contains fifty-one tanks (2 stabilized, 49 yet to be closed) distributed between two liquid radioactive waste storage facilities at SRS containing carbon steel underground tanks with storage capacities ranging from 2,800,000 to 4,900,000 liters. Treatment of the liquid waste from these tanks is essential both to closing older tanks and to maintaining space needed to treat the waste that is eventually vitrified or disposed of onsite. Section 3116 of the Ronald W. Reagan National Defense Authorization Act of Fiscal Year 2005 (NDAA) provides the Secretary of Energy, in consultation with the Nuclear Regulatory Commission (NRC), a methodology to determine that certain waste resulting from prior reprocessing of spent nuclear fuel are not high-level radioactive waste if it can be demonstrated that the waste meets the criteria set forth in Section 3116(a) of the NDAA. The Secretary of Energy, in consultation with the NRC, signed a determination in January 2006, pursuant to Section 3116(a) of the NDAA, for salt waste disposal at the SRS Saltstone Disposal Facility. This determination is based, in part, on the Basis for Section 3116 Determination for Salt Waste Disposal at the Savannah River Site and supporting references, a document that describes the planned methods of liquid waste treatment and the resulting waste streams. The document provides descriptions of the proposed methods for processing salt waste, dividing them into 'Interim Salt Processing' and later processing through the Salt Waste Processing Facility (SWPF). Interim Salt Processing is separated into Deliquification, Dissolution, and Adjustment (DDA) and Actinide Removal Process/Caustic Side Solvent Extraction Unit (ARP/MCU). The Waste Determination was signed

Establishment of new disposal capacity for the Savannah River Plant

International Nuclear Information System (INIS)

Albenesius, E.L.; Wilhite, E.L.

1987-01-01

Two new low-level waste (LLW) disposal sites for decontaminated salt solidified with cement and fly ash (saltstone) and for conventional solid LLW are planned for SRP in the next several years. An above-ground vault disposal system for saltstone was designed to minimize impact on the environment by controlling permeability and diffusivity of the waste form and concrete liner. The experimental program leading to the engineered disposal system included formulation studies, multiple approaches to measurement of permeability and diffusivity, extensive mathematical modeling, and large-scale lysimeter tests to validate model projections. The overall study is an example of the systems approach to disposal site design to achieve a predetermined performance objective. The same systems approach is being used to develop alternative designs for disposal of conventional LLW at the Savannah River Plant. 14 figures

Environmental Restoration Disposal Facility waste acceptance criteria. Revision 1

International Nuclear Information System (INIS)

Corriveau, C.E.

1996-01-01

The Environmental Restoration Disposal Facility (ERDF) is designed to be an isolation structure for low-level radioactive remediation waste, chemically contaminated remediation waste, and remediation waste that contains both chemical and radioactive constituents (i.e., mixed remediation waste) produced during environmental remediation of Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) past-practice units at the Hanford Site. Remedial action wastes, which will become a structural component of the ERDF, include bulk soil, demolition debris, and miscellaneous wastes from burial grounds. These wastes may originate from CERCLA past-practice sites (i.e., operable units) in the 100 Areas, the 200 Areas, and the 300 Area of the Hanford Site

An approach for sampling solid heterogeneous waste at the Hanford Site waste receiving and processing and solid waste projects

International Nuclear Information System (INIS)

Sexton, R.A.

1993-03-01

This paper addresses the problem of obtaining meaningful data from samples of solid heterogeneous waste while maintaining sample rates as low as practical. The Waste Receiving and Processing Facility, Module 1, at the Hanford Site in south-central Washington State will process mostly heterogeneous solid wastes. The presence of hazardous materials is documented for some packages and unknown for others. Waste characterization is needed to segregate the waste, meet waste acceptance and shipping requirements, and meet facility permitting requirements. Sampling and analysis are expensive, and no amount of sampling will produce absolute certainty of waste contents. A sampling strategy is proposed that provides acceptable confidence with achievable sampling rates

Statistical experimental design for saltstone mixtures

International Nuclear Information System (INIS)

Harris, S.P.; Postles, R.L.

1992-01-01

The authors used a mixture experimental design for determining a window of operability for a process at the U.S. Department of Energy, Savannah River Site, Defense Waste Processing Facility (DWPF). The high-level radioactive waste at the Savannah River Site is stored in large underground carbon steel tanks. The waste consists of a supernate layer and a sludge layer. Cesium-137 will be removed from the supernate by precipitation and filtration. After further processing, the supernate layer will be fixed as a grout for disposal in concrete vaults. The remaining precipitate will be processed at the DWPF with treated waste tank sludge and glass-making chemicals into borosilicate glass. The leach-rate properties of the supernate grout formed from various mixes of solidified coefficients for NO 3 and chromium were used as a measure of leach rate. Various mixes of cement, Ca(OH) 2 , salt, slag, and fly ash were used. These constituents comprise the whole mix. Thus, a mixture experimental design was used. The regression procedure (PROC REG) in SAS was used to produce analysis of variance (ANOVA) statistics. In addition, detailed model diagnostics are readily available for identifying suspicious observations. For convenience, trillinear contour (TLC) plots, a standard graphics tool for examining mixture response surfaces, of the fitted model were produced using ECHIP

Immobilization and Waste Form Product Acceptance for Low Level and TRU Waste Forms

International Nuclear Information System (INIS)

Holtzscheiter, E.W.; Harbour, J.R.

1998-05-01

The Tanks Focus Area is supporting technology development in immobilization of both High Level (HLW) and Low Level (LLW) radioactive wastes. The HLW process development at Hanford and Idaho is patterned closely after that of the Savannah River (Defense Waste Processing Facility) and West Valley Sites (West Valley Demonstration Project). However, the development and options open to addressing Low Level Waste are diverse and often site specific. To start, it is important to understand the breadth of Low Level Wastes categories

Disposal of decontaminated salts at the Savannah River Plant by solidification and burial

International Nuclear Information System (INIS)

Dukes, M.D.; Wolf, H.C.; Langton, C.A.

1983-01-01

The current plan for disposal of waste salt at the Savannah River Plant (SRP) is to immobilize the decontaminated salt solution by mixing with cement and SRP soil, and bury the resulting grout (saltstone) in a landfill. The grout which contains 37.8 wt % salt solution, 22.8 wt % Portland I-P cement, and 39.2 wt % SRP soil, was specially formulated to have a low permeability ( -10 cm/sec). This material will be mixed and placed in trenches. After setting, the saltstone will be covered with a clay cap, and an overburden of compacted native soil will be replaced. 6 references

Bases, Assumptions, and Results of the Flowsheet Calculations for the Decision Phase Salt Disposition Alternatives

Energy Technology Data Exchange (ETDEWEB)

Elder, H.H.

2001-07-11

The HLW salt waste (salt cake and supernate) now stored at the SRS must be treated to remove insoluble sludge solids and reduce the soluble concentration of radioactive cesium radioactive strontium and transuranic contaminants (principally Pu and Np). These treatments will enable the salt solution to be processed for disposal as saltstone, a solid low-level waste.

Waste Analysis Plan for the Waste Receiving and Processing (WRAP) Facility

International Nuclear Information System (INIS)

TRINER, G.C.

1999-01-01

The purpose of this waste analysis plan (WAP) is to document the waste acceptance process, sampling methodologies, analytical techniques, and overall processes that are undertaken for dangerous, mixed, and radioactive waste accepted for confirmation, nondestructive examination (NDE) and nondestructive assay (NDA), repackaging, certification, and/or storage at the Waste Receiving and Processing Facility (WRAP). Mixed and/or radioactive waste is treated at WRAP. WRAP is located in the 200 West Area of the Hanford Facility, Richland, Washington. Because dangerous waste does not include source, special nuclear, and by-product material components of mixed waste, radionuclides are not within the scope of this documentation. The information on radionuclides is provided only for general knowledge

From waste to technology. An environmental acceptable technology

International Nuclear Information System (INIS)

Tabasaran, O.

2001-01-01

Modern thermal waste treatment has the task to produce deposit capable inert products and is able to reduce contaminants emitted to the environment. In addition waste to energy plants enable the recovery of energy by a maximum amount of environmental protection even in the comparison with modern power stations or other industrial plants. Over that, regarding modern technical solutions, today's waste to energy plants can no more be turned on financial reasons [it

Assessment of LANL transuranic mixed waste management documentation

International Nuclear Information System (INIS)

Davis, K.D.; Hoevemeyer, S.S.; McCance, C.H.; Jennrich, E.A.; Lund, D.M.

1991-04-01

The objective of this report is to present findings from the evaluation of the Los Alamos National Laboratory (LANL) TRU Mixed Waste Acceptance Criteria to determine its compliance with applicable DOE requirements. The driving requirements for s TRU Mixed Waste Acceptance Criteria are essentially those contained in the ''TRU Waste Acceptance Criteria for the Waste Isolation Pilot Plant'' or WIPP WAC (DOE Report WIPP-DOE-069), 40 CFR 261-270, and DOE Order 5820.2A (Radioactive Waste Management), specifically Chapter II which is entitled ''Management of Transuranic Waste''. The primary purpose of the LANL WAC is the establishment of those criteria that must be met by generators of TRU mixed waste before such waste can be accepted by the Waste Management Group. An annotated outline of a genetic TRU mixed waste acceptance criteria document was prepared from those requirements contained in the WIPP WAC, 40 CFR 261-270, and 5820.2A, and is based solely upon those requirements

Preliminary Hanford Waste Vitrification Plan Waste Form Qualification Plan

International Nuclear Information System (INIS)

Nelson, J.L.

1987-09-01

This Waste Form Qualification Plan describes the waste form qualification activities that will be followed during the design and operation of the Hanford Waste Vitrification Plant to ensure that the vitrified Hanford defense high-level wastes will meet the acceptance requirements of the candidate geologic repositories for nuclear waste. This plan is based on the defense waste processing facility requirements. The content of this plan is based on the assumption that the Hanford Waste Vitrification Plant high-level waste form will be disposed of in one of the geologic repository projects. Proposed legislation currently under consideration by Congress may change or delay the repository site selection process. The impacts of this change will be assessed as details of the new legislation become available. The Plan describes activities, schedules, and programmatic interfaces. The Waste Form Qualification Plan is updated regularly to incorporate Hanford Waste Vitrification Plant-specific waste acceptance requirements and to serve as a controlled baseline plan from which changes in related programs can be incorporated. 10 refs., 5 figs., 5 tabs

Can marketing increase the legitimacy and acceptance of nuclear waste management?

International Nuclear Information System (INIS)

Kurki, Osmo

1999-01-01

Full text: In Finland the most important decision concerning the nuclear waste management is a decision in principle. The decision in principle of the Council of State is needed for the final disposal of nuclear waste. Decision is made by the Finnish Government and then approved or rejected by the Finnish Parliament. According to Posiva's plan we are going to apply for the decision this spring. If we succeed, a site for the final disposal will be chosen and all investigations concentrated to the selected area. For now we have four candidate areas. The decision will be political - it is a fact. To be able to get the decision in principle approved requires that at least half of the 200 representatives of the parliament vote yes. To get acceptance both in national and municipal levels is an exceptionally hard task. AR possible communication and marketing mixes are needed in order to reach the goal. During Posiva's brief history three national marketing campaigns have been made. The results have been encouraging. Swedish nuclear transport vessel M/S Sigyn has played an essential part in an our national marketing campaigns. An exhibition onboard has visited several harbours in Finland. This year Posiva's marketing event was divided in three parts - Advertising campaign, - Sigyn's visit to Helsinki, the capital city of Finland, - Expert seminars. All parts were linked to each other. We launched the campaign three weeks before Sigyn's visit. Main message was: 'Not everyone has to be highly precise with their facts. We must.' The main goal of the advertisements published in national magazines and newspapers was to increase the publicity of Posiva and the Finnish solution for the nuclear waste management. At the same time we hoped to create public debate and open discussion, which is always needed in a democratic society before important national decisions. We also used movie advertising. A humoristic approach, which is always a demanding genre, was chosen. The results of

Social acceptance process model for ensuring the high-level radioactive waste disposal site

International Nuclear Information System (INIS)

Wada, Ryutaro; Tanaka, Satoru; Nagasaki, Shinya

2009-01-01

Generally speaking, a vast, advanced and unfamiliar science and technology are unacceptable to the public for fear of their unknown nature. Here, the social acceptance process model was examined on the basis of the analysis of the cause phenomenon and numerical grounds, by referring to the problems on the application of literature documentation for location examination of a high-level radioactive waste disposal site in Toyo town in Kochi Pref. in April 2007. In analyzing the Toyo town case, we have found a possibility that the majority of local residents knew very little about the object opposed by the fringe route processing. To ensure a healthy decision making by the public, it is vital to convey fundamental information using sufficient wide-area PR media before the issue becomes actual. After the issue becomes actual, dialog with residents through a careful technology assessment is indispensable. The authors focus attention on the decision-making process of human beings from the social and psychological viewpoints, and point out that it is desirable for promoting social acceptance by adopting two approaches: a direct approach aiming at better intelligibility for the different resident layers and a deductive approach in technological essence. (author)

Infectious waste feed system

Science.gov (United States)

Coulthard, E. James

1994-01-01

An infectious waste feed system for comminuting infectious waste and feeding the comminuted waste to a combustor automatically without the need for human intervention. The system includes a receptacle for accepting waste materials. Preferably, the receptacle includes a first and second compartment and a means for sealing the first and second compartments from the atmosphere. A shredder is disposed to comminute waste materials accepted in the receptacle to a predetermined size. A trough is disposed to receive the comminuted waste materials from the shredder. A feeding means is disposed within the trough and is movable in a first and second direction for feeding the comminuted waste materials to a combustor.

Methods for maintaining a record of waste packages during waste processing and storage

International Nuclear Information System (INIS)

2005-01-01

During processing, radioactive waste is converted into waste packages, and then sent for storage and ultimately for disposal. A principal condition for acceptance of a waste package is its full compliance with waste acceptance criteria for disposal or storage. These criteria define the radiological, mechanical, physical, chemical and biological properties of radioactive waste that can, in principle, be changed during waste processing. To declare compliance of a waste package with waste acceptance criteria, a system for generating and maintaining records should be established to record and track all relevant information, from raw waste characteristics, through changes related to waste processing, to final checking and verification of waste package parameters. In parallel, records on processing technology and the operational parameters of technological facilities should adhere to established and approved quality assurance systems. A records system for waste management should be in place, defining the data to be collected and stored at each step of waste processing and using a reliable selection process carried over into the individual steps of the waste processing flow stream. The waste management records system must at the same time ensure selection and maintenance of all the main information, not only providing evidence of compliance of waste package parameters with waste acceptance criteria but also serving as an information source in the case of any future operations involving the stored or disposed waste. Records generated during waste processing are a constituent part of the more complex system of waste management record keeping, covering the entire life cycle of radioactive waste from generation to disposal and even the post-closure period of a disposal facility. The IAEA is systematically working on the preparation of a set of publications to assist its Member States in the development and implementation of such a system. This report covers all the principal

Neptunium sorption and co-precipitation of strontium in simulated DWPF salt solution

International Nuclear Information System (INIS)

McIntyre, P.F.; Orebaugh, E.G.; King, C.M.

1988-01-01

Batch experiments performed using crushed slag saltstone (∼40 mesh) removed >80% of 237 Np from simulated Defense Waste Processing Facility (DWPF) salt solution. The concentration of 237 Np (110 pCi/ml) used was 1000x greater than levels in actual DWPF solutions. Neptunium-239 was used as a tracer and was formed by neutron activation of uranyl nitrate. Results showed that small amounts of crushed saltstone (as little as 0.05 grams), removed >80% of neptunium from 15 ml of simulated DWPF solution after several hours equilibration. The neptunium is sorbed on insoluble carbonates formed in and on the saltstone matrix. Further testing showed that addition of 0.01 and 0.10 ml of 1 molar Ca +2 (ie. Ca (NO 3 ) 2 , CaCl 2 ) into 15 ml of simulated DWPF solution yielded a white carbonate precipitate which also removed >80% of the neptunium after 1 hour equilibration. Further experiments were performed to determine the effectiveness of this procedure to co-precipitate strontium

Development of new waste form for treatment and disposal of concentrated liquid radioactive waste

International Nuclear Information System (INIS)

Kwak, Kyung Kil; Ji, Young Yong

2010-12-01

The radioactive waste form should be meet the waste acceptance criteria of national regulation and disposal site specification. We carried out a characterization of rad waste form, especially the characteristics of radioactivity, mechanical and physical-chemical properties in various rad waste forms. But asphalt products is not acceptable waste form at disposal site. Thus we are change the product materials. We select the development of the new process or new materials. The asphalt process is treatment of concentrated liquid and spent-resin and that we decide the Development of new waste form for treatment and disposal of concentrated liquid radioactive waste

Hanford Immobilized Low-Activity Waste Product Acceptance Test Plan

International Nuclear Information System (INIS)

Peeler, D.

1999-01-01

'The Hanford Site has been used to produce nuclear materials for the U.S. Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during Pu production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The DOE is proceeding with an approach to privatize the treatment and immobilization of Handord''s LAW and HLW.'

Waste Analysis Plan for 241-Z

International Nuclear Information System (INIS)

HIRZEL, D.R.

2000-01-01

The 241-2 waste tanks are used to store, treat, and transfer waste to Tank Farms. The sampling requirements are established to identify the composition of the tank waste. The primary goal is to meet the Tank Farms acceptance criteria. Tank Farms will not accept waste without extensive characterization sample data. Process and lab wastes are sampled for suitability prior to routing to Tk-D8. The samples are helpful in tracking the amount of chemical constituents to determine treatment and are required to maintain Pu inventory and criticality prevention limitations. Likewise, the waste is sampled prior to inter-tank transfers. The revised Waste Analysis Plan for 241-2 (WAP) contains current facility, process and waste descriptions. The WAP lists the Double Shell Tank (DST) system acceptance criteria, sampling parameters and required analyses. The characterization data on historical process wastes was deleted. A section on the Tank Farms waste approval procedural process was added to describe the steps necessary and documentation required to transfer waste to the DST system. Failure to collect proper samples will result in Tank Farms' refusal to accept PFP waste until proper sampling conditions are met. This will use up unnecessary time and resources but not place the plant in a hazardous position

Hanford Immobilized Low-Activity Waste Product Acceptance Test Plan

Energy Technology Data Exchange (ETDEWEB)

Peeler, D.

1999-06-22

'The Hanford Site has been used to produce nuclear materials for the U.S. Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during Pu production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The DOE is proceeding with an approach to privatize the treatment and immobilization of Handord''s LAW and HLW.'

PIME '98, proposal for opening contribution: Nuclear waste

International Nuclear Information System (INIS)

Raurnolin, Heikki

1998-01-01

Full text: Would a debate about an international nuclear waste repository help us win greater public acceptance for our disposal plans? My opening points will be: - International nuclear waste repositories can be accepted by the public only after the acceptance of national repositories. If there are no accepted national plans or existing national repositories, nobody is accepting any international repository in his or her own country; - The focus of gaining public acceptance should therefore be on the national programmes and on the technology itself, i.e. 'Deep disposal is a safe solution independent on the type of rock formations, crystalline, salt, clay etc.'; - The Finnish situation is quite clear. Our people are rather confident on the stability of our old crystalline granite bedrock. Finnish politicians and ordinary people are very much against accepting high-level waste or spent nuclear fuel of foreign origin to be disposed of in Finland. This was one of the reasons why the Finnish Nuclear Act was amended before Finland joined to EU, so that the import and export of nuclear waste are forbidden; - Our site selection programme in Finland is in a very sensitive phase. The Government has just confirmed the target, site selection at the end of year 2000, and the statutory Environmental Impact Assesment process has just been initiated in four candidate sites. Certain opponents try to frighten people by claiming that accepting the site and the deep disposal of our domestic waste means also definitely accepting the same for foreign waste, in any case for any nuclear waste from other EU countries; - So, all news on discussion about international nuclear waste repositories will create more suspicions against the Finnish nuclear authorities, waste company and utilities. Summary: The answer is no, the debate about international nuclear waste repository does not help us to win greater public acceptance for our disposal plans. (author)

Aspects on the acceptance of waste for disposal in SFR

International Nuclear Information System (INIS)

Torstenfelt, Boerje

2006-01-01

When licensing a final repository for radioactive waste certain assumptions have to be made concerning the waste. These assumptions cover radionuclide inventory and nonradiological materials and its physical and chemical impact on the waste, the repository and on the environment. Development of new waste treatment systems and waste packages at the waste producer site aim at finding solutions and products that can be stored, transported and disposed of safely and are economically sound. This paper discusses some aspects concerning development of new or modified waste products. It highlights the importance of analysing the whole sequence in treatment, handling and disposing the waste. The process should be to find an optimal solution for the whole system, considering the fact that what is best in one step it not necessary best for the whole system, including the post closure issues. (author)

Goals for nuclear waste management

International Nuclear Information System (INIS)

Watson, R.A.

1978-01-01

Establishing a publicly, politically, economically, and technologically acceptable waste management system for the fuel cycle is a necessary condition for accepting the nuclear program as a national energy option. Findings are given on the technology, politics, economics, morality, aesthetics, and societal impact of waste management. Proposed goals are outlined for the regulation of waste management

Assessment of national systems for obtaining local acceptance of nuclear waste management siting activities

International Nuclear Information System (INIS)

Paige, H.W.; Owens, J.E.

1984-01-01

On behalf of the United States Department of Energy (DOE), International Energy Associates Limited (IEAL) of Washington, D.C. has conducted surveys and analyses of fourteen countries' plans and approaches for dealing with the problems of obtaining local siting acceptance for nuclear waste management facilities. It was determined that the following elements of the formal systems generally facilitate and/or expedite waste management siting decisions: (1) a clear-cut pro-nuclear power position on the part of the government; (2) a willingness on the part of the central government to exert (with prudence and restraint) its pre-emptive rights in nuclear matters; (3) political structures in which the heads of regional or provincial governments are appointed by the central government; (4) national laws that link reactor licensing with a detailed plan for waste management; (5) an established and stable policy with regard to reprocessing. In contrast, it was determined that the following elements of the formal system generally hinder waste management siting activities: (1) historically strong local land used veto laws; (2) the use of national referenda for making nuclear decisions; (3) requirements for public hearings. The informal approaches fall into the following five categories: (1) political: e.g. assertion of will by political leaders, activities to enlist support of local politicians, activities to broaden involvement in decision-making; (2) economic: e.g. emphasis on normal benefits, provision for additional economic benefits; (3) siting: e.g. at or near existing nuclear facilities, on government or utility property, at multiple locations to spread the political burden; (4) timing: e.g. decoupling drilling activities from ultimate repository site decision, deliberate deferral to (long-range) future; (5) education: e.g. creation of special government programmes, enlisting of media support

TRU Waste Sampling Program: Volume I. Waste characterization

International Nuclear Information System (INIS)

Clements, T.L. Jr.; Kudera, D.E.

1985-09-01

Volume I of the TRU Waste Sampling Program report presents the waste characterization information obtained from sampling and characterizing various aged transuranic waste retrieved from storage at the Idaho National Engineering Laboratory and the Los Alamos National Laboratory. The data contained in this report include the results of gas sampling and gas generation, radiographic examinations, waste visual examination results, and waste compliance with the Waste Isolation Pilot Plant-Waste Acceptance Criteria (WIPP-WAC). A separate report, Volume II, contains data from the gas generation studies

Bridging the Gap between Social Acceptance and Ethical Acceptability.

Science.gov (United States)

Taebi, Behnam

2017-10-01

New technology brings great benefits, but it can also create new and significant risks. When evaluating those risks in policymaking, there is a tendency to focus on social acceptance. By solely focusing on social acceptance, we could, however, overlook important ethical aspects of technological risk, particularly when we evaluate technologies with transnational and intergenerational risks. I argue that good governance of risky technology requires analyzing both social acceptance and ethical acceptability. Conceptually, these two notions are mostly complementary. Social acceptance studies are not capable of sufficiently capturing all the morally relevant features of risky technologies; ethical analyses do not typically include stakeholders' opinions, and they therefore lack the relevant empirical input for a thorough ethical evaluation. Only when carried out in conjunction are these two types of analysis relevant to national and international governance of risky technology. I discuss the Rawlsian wide reflective equilibrium as a method for marrying social acceptance and ethical acceptability. Although the rationale of my argument is broadly applicable, I will examine the case of multinational nuclear waste repositories in particular. This example will show how ethical issues may be overlooked if we focus only on social acceptance, and will provide a test case for demonstrating how the wide reflective equilibrium can help to bridge the proverbial acceptance-acceptability gap. © 2016 The Authors Risk Analysis published by Wiley Periodicals, Inc. on behalf of Society for Risk Analysis.

Towards optimization of nuclear waste glass: Constraints, property models, and waste loading

International Nuclear Information System (INIS)

Hrma, P.

1994-04-01

Vitrification of both low- and high-level wastes from 177 tanks at Hanford poses a great challenge to glass makers, whose task is to formulate a system of glasses that are acceptable to the federal repository for disposal. The enormous quantity of the waste requires a glass product of the lowest possible volume. The incomplete knowledge of waste composition, its variability, and lack of an appropriate vitrification technology further complicates this difficult task. A simple relationship between the waste loading and the waste glass volume is presented and applied to the predominantly refractory (usually high-activity) and predominantly alkaline (usually low-activity) waste types. Three factors that limit waste loading are discussed, namely product acceptability, melter processing, and model validity. Glass formulation and optimization problems are identified and a broader approach to uncertainties is suggested

Assessment of national systems for obtaining local siting acceptance of nuclear-waste-management facilities (1981). Final report

International Nuclear Information System (INIS)

1981-01-01

There is a rich mixture of formal and informal approaches being used in our sister nuclear democracies in their attempts to deal with the difficulties in obtaining local siting acceptance of national waste management facilities. Some of these are meeting with a degree of success not yet achieved in the US. Although this survey documents and assesses many of these approaches, the scope of the study did not include an assessment of their relevance to common problems in the US. It would appear that in addition to a periodic updating of the approaches and progress of other countries in dealing with the siting of nuclear waste facilities, an assessment of the applicability of the more successful of these approaches to the US political system could make good use of the information developed in the preparation of this report

Deconstructing the Nature of Safety and Acceptability in Nuclear Fuel Waste Management, Canada

International Nuclear Information System (INIS)

Murphy, Brenda L.

2003-01-01

Since the Hare report was released in 1977, Canada's nuclear fuel waste management policies have been focused on one option, deep geologic disposal. However, since new legislation, called the Nuclear Fuel Waste Management Act came into force on November 15, 2002, the newly established Nuclear Waste Management Organisation has been mandated to review, over the next three years, not only deep geologic disposal but also to consider a suite of other options including reactor-site and centralised storage, and both above and below ground options. At this point in time, the process for the study and the criteria by which it will be evaluated are still quite unclear. The need for this new approach to NFW management in Canada was highlighted by an Environmental Assessment Panel (EA Panel) report in 1998. This EA Panel reviewed the 1994 environmental impact statement submitted by Atomic Energy Canada Limited (AECL) regarding the concept of deep geologic disposal in the Canadian Shield (Disposal Concept EA). The EA Panel's key conclusion was: From a technical perspective, safety of the AECL concept has been on balance adequately demonstrated for a conceptual stage of development, but from a social perspective, it has not. As it stands, the AECL concept for deep geological disposal has not been demonstrated to have broad public support. The concept in its current form does not have the required level of acceptability to be adopted as Canada's approach for managing nuclear fuel wastes. AECL's study consisted of a review of the deep geologic disposal concept, since no site has yet been selected. It is generally understood that this type of disposal facility, if eventually sited, will be located on the Ontario portion of the Canadian Shield because Ontario power generators own and manage about 90% of Canada's NFW. This key EA Panel conclusion is predicated on the use of two key decision-making criteria - safety and acceptability. This paper reviews and assesses the specific ways

Acceptance test procedure for core sample trucks

International Nuclear Information System (INIS)

Smalley, J.L.

1995-01-01

The purpose of this Acceptance Test Procedure is to provide instruction and documentation for acceptance testing of the rotary mode core sample trucks, HO-68K-4600 and HO-68K-4647. The rotary mode core sample trucks were based upon the design of the second core sample truck (HO-68K-4345) which was constructed to implement rotary mode sampling of the waste tanks at Hanford. Acceptance testing of the rotary mode core sample trucks will verify that the design requirements have been met. All testing will be non-radioactive and stand-in materials shall be used to simulate waste tank conditions. Compressed air will be substituted for nitrogen during the majority of testing, with nitrogen being used only for flow characterization

Hanford Site Hazardous waste determination report for transuranic debris waste streams NPFPDL2A

International Nuclear Information System (INIS)

WINTERHALDER, J.A.

1999-01-01

This hazardous waste determination report (Report) describes the process and information used on the Hanford Site to determine that waste stream number NPFPDLZA, consisting of 30 containers of contact-handled transuranic debris waste, is not hazardous waste regulated by the Resource Conservation and Recovery Act (RCRA) or the New Mexico Hazardous Waste Act. For a waste to be hazardous under these statutes, the waste either must be specifically listed as a hazardous waste, or exhibit one or more of the characteristics of a hazardous waste, Le., ignitability, corrosivity, reactivity, or toxicity. Waste stream NPFPDLZA was generated, packaged, and placed into storage between 1993 and 1997. Extensive knowledge of the waste generating process, facility operational history, and administrative controls and operating procedures in effect at the time of generation, supported the initial nonhazardous waste determination. Because of the extent and reliability of information pertaining to this waste type, and the total volume of waste in the debris matrix parameter category, the Hanford Site is focusing initial efforts on this and similar waste streams for the first shipment to the Waste Isolation Pilot Plant (WIPP). RCRA regulations authorize hazardous waste determinations to be made either by using approved sampling and analysis methods or by applying knowledge of the waste in light of the materials or the process(es) used. This latter approach typically is referred to as process knowledge. The Transuranic Waste Characterization Quality Assurance Program Plan (CAO-94-1010) for WIPP refers to acceptable knowledge in essentially the same terms; acceptable knowledge as used throughout this Report is synonymous with the term process knowledge. The 30 containers addressed in this Report were characterized by the following methods: Acceptable knowledge; Nondestructive examination using real-time radiography; Visual examination; and Headspace gas sampling and analysis. The initial

Hanford Site Hazardous waste determination report for transuranic debris waste streams NPFPDL2A

Energy Technology Data Exchange (ETDEWEB)

WINTERHALDER, J.A.

1999-09-29

This hazardous waste determination report (Report) describes the process and information used on the Hanford Site to determine that waste stream number NPFPDLZA, consisting of 30 containers of contact-handled transuranic debris waste, is not hazardous waste regulated by the Resource Conservation and Recovery Act (RCRA) or the New Mexico Hazardous Waste Act. For a waste to be hazardous under these statutes, the waste either must be specifically listed as a hazardous waste, or exhibit one or more of the characteristics of a hazardous waste, Le., ignitability, corrosivity, reactivity, or toxicity. Waste stream NPFPDLZA was generated, packaged, and placed into storage between 1993 and 1997. Extensive knowledge of the waste generating process, facility operational history, and administrative controls and operating procedures in effect at the time of generation, supported the initial nonhazardous waste determination. Because of the extent and reliability of information pertaining to this waste type, and the total volume of waste in the debris matrix parameter category, the Hanford Site is focusing initial efforts on this and similar waste streams for the first shipment to the Waste Isolation Pilot Plant (WIPP). RCRA regulations authorize hazardous waste determinations to be made either by using approved sampling and analysis methods or by applying knowledge of the waste in light of the materials or the process(es) used. This latter approach typically is referred to as process knowledge. The Transuranic Waste Characterization Quality Assurance Program Plan (CAO-94-1010) for WIPP refers to acceptable knowledge in essentially the same terms; acceptable knowledge as used throughout this Report is synonymous with the term process knowledge. The 30 containers addressed in this Report were characterized by the following methods: Acceptable knowledge; Nondestructive examination using real-time radiography; Visual examination; and Headspace gas sampling and analysis. The initial

Is a Finnish municipality going to accept the radioactive wastes

International Nuclear Information System (INIS)

Harmaajaervi, I.

1997-01-01

The article reviews the methods and results of the questionnaire study collecting information on the public attitudes with regard to radioactive waste disposal in Finland. The study started in December 1996 and concentrated on the population living in the areas of radioactive waste disposal site investigations, Eurajoki, Aeaenekoski and Kuhmo. The items of the questionnaire asked about the impacts of the radioactive waste disposal on the environment, and also on public health and general safety. The results addressed the importance of the relevant and comprehensive information of the radioactive waste management to the public. The study is a part of the Publicly administrated nuclear waste management programme (JYT2) in Finland

Mixed Waste Focus Area - Waste form initiative

International Nuclear Information System (INIS)

Nakaoka, R.; Waters, R.; Pohl, P.; Roach, J.

1998-01-01

The mission of the US Department of Energy's (DOE) Mixed Waste Focus Area (MWFA) is to provide acceptable technologies that enable implementation of mixed waste treatment systems which are developed in partnership with end-users, stakeholders, tribal governments, and regulators. To accomplish this mission, a technical baseline was established in 1996 and revised in 1997. The technical baseline forms the basis for determining which technology development activities will be supported by the MWFA. The primary attribute of the technical baseline is a set of prioritized technical deficiencies or roadblocks related to implementation of mixed waste treatment systems. The Waste Form Initiative (WFI) was established to address an identified technical deficiency related to waste form performance. The primary goal of the WFI was to ensure that the mixed low-level waste (MLLW) treatment technologies being developed, currently used, or planned for use by DOE would produce final waste forms that meet the waste acceptance criteria (WAC) of the existing and/or planned MLLW disposal facilities. The WFI was limited to an evaluation of the disposal requirements for the radioactive component of MLLW. Disposal requirements for the hazardous component are dictated by the Resource Conservation and Recovery Act (RCRA), and were not addressed. This paper summarizes the technical basis, strategy, and results of the activities performed as part of the WFI

Guidelines for generators of hazardous chemical waste at LBL and Guidelines for generators of radioactive and mixed waste at LBL

International Nuclear Information System (INIS)

1991-07-01

The purpose of this document is to provide the acceptance criteria for the transfer of hazardous chemical, radioactive, and mixed waste to Lawrence Berkeley Laboratory's (LBL) Hazardous Waste Handling Facility (HWHF). These guidelines describe how a generator of wastes can meet LBL's acceptance criteria for hazardous chemical, radioactive, and mixed waste. 9 figs

Variation in information use in the evaluation of the social acceptability of nuclear waste management options

International Nuclear Information System (INIS)

Allen, P.T.

1991-01-01

This article considers how technical information is used in forming judgements about such complex technological phenomena as nuclear waste management. Public understanding and acceptance of innovations is also considered, particularly where health and safety issues are involved. Education of the public is not a total solution. Many problems include political, social or ethical aspects which impinge on how they are judged by non-scientists. In addition, the vested interests of various parties must be weighed. A field-based experiment is described which is designed to illustrate the difference between ''normal'' and ''normative'' evaluation procedures used by both technically experienced and lay participants. (UK)

Variability Of KD Values In Cementitious Materials And Sediments

International Nuclear Information System (INIS)

Almond, P.; Kaplan, D.; Shine, E.

2012-01-01

Measured distribution coefficients (K d values) for environmental contaminants provide input data for performance assessments (PA) that evaluate physical and chemical phenomena for release of radionuclides from wasteforms, degradation of engineered components and subsequent transport of radionuclides through environmental media. Research efforts at SRNL to study the effects of formulation and curing variability on the physiochemical properties of the saltstone wasteform produced at the Saltstone Disposal Facility (SDF) are ongoing and provide information for the PA and Saltstone Operations. Furthermore, the range and distribution of plutonium K d values in soils is not known. Knowledge of these parameters is needed to provide guidance for stochastic modeling in the PA. Under the current SRS liquid waste processing system, supernate from F and H Tank Farm tanks is processed to remove actinides and fission products, resulting in a low-curie Decontaminated Salt Solution (DSS). At the Saltstone Production Facility (SPF), DSS is mixed with premix, comprised of blast furnace slag (BFS), Class F fly ash (FA), and portland cement (OPC) to form a grout mixture. The fresh grout is subsequently placed in SDF vaults where it cures through hydration reactions to produce saltstone, a hardened monolithic waste form. Variation in saltstone composition and cure conditions of grout can affect the saltstone's physiochemical properties. Variations in properties may originate from variables in DSS, premix, and water to premix ratio, grout mixing, placing, and curing conditions including time and temperature (Harbour et al. 2007; Harbour et al. 2009). There are no previous studies reported in the literature regarding the range and distribution of K d values in cementitious materials. Presently, the Savannah River Site (SRS) estimate ranges and distributions of K d values based on measurements of K d values made in sandy SRS sediments (Kaplan 2010). The actual cementitious material K d

Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project

International Nuclear Information System (INIS)

2016-05-01

Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants

Assessment of LANL hazardous waste management documentation

International Nuclear Information System (INIS)

Davis, K.D.; Hoevemeyer, S.S.; Stirrup, T.S.; Jennrich, E.A.; Lund, D.M.

1991-04-01

The objective of this report is to present findings from evaluating the Los Alamos National Laboratory (LANL) ''Hazardous Waste Acceptance Criteria Receipt at TA-54, Area L'' to determine if it meets applicable DOE requirements. The guidelines and requirements for the establishment of a Hazardous Waste Acceptance Criteria (HW-WAC) are found in 40 CFR 260 to 270 and DOE Order 5820.2A. Neither set of requirements specifically require a WAC for the management of hazardous waste; however, the use of such documentation is logical and is consistent with the approach required for the management of radioactive waste. The primary purpose of a HW-WAC is to provide generators and waste management with established criteria that must be met before hazardous waste can be acceptable for treatment, storage and/or disposal. An annotated outline for a generic waste acceptance criteria was developed based on the requirements of 40 CFR 260 to 270 and DOE Order 5820.2A. The outline contains only requirements for hazardous waste, it does not address the radiological components of low-level mixed waste. The outline generated from the regulations was used for comparison to the LANL WAC For Chemical and Low-level Mixed Waste Receipt at TA-54, Area L. The major elements that should be addressed by a hazardous waste WAC were determined to be as follows: Waste Package/Container Requirements, Waste Forms, Land Disposal Restrictions, and Data Package-Certification ampersand Documentation

Waste Characterization Methods

Energy Technology Data Exchange (ETDEWEB)

Vigil-Holterman, Luciana R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Naranjo, Felicia Danielle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-02-02

This report discusses ways to classify waste as outlined by LANL. Waste Generators must make a waste determination and characterize regulated waste by appropriate analytical testing or use of acceptable knowledge (AK). Use of AK for characterization requires several source documents. Waste characterization documentation must be accurate, sufficient, and current (i.e., updated); relevant and traceable to the waste stream’s generation, characterization, and management; and not merely a list of information sources.

Waste Characterization Methods

International Nuclear Information System (INIS)

Vigil-Holterman, Luciana R.; Naranjo, Felicia Danielle

2016-01-01

This report discusses ways to classify waste as outlined by LANL. Waste Generators must make a waste determination and characterize regulated waste by appropriate analytical testing or use of acceptable knowledge (AK). Use of AK for characterization requires several source documents. Waste characterization documentation must be accurate, sufficient, and current (i.e., updated); relevant and traceable to the waste stream's generation, characterization, and management; and not merely a list of information sources.

Evaluation Of The Impact Of The Defense Waste Processing Facility (DWPF) Laboratory Germanium Oxide Use On Recycle Transfers To The H-Tank Farm

International Nuclear Information System (INIS)

Jantzen, C.; Laurinat, J.

2011-01-01

goal of 400 canisters. Since no Waste Acceptance Criteria (WAC) exists for germanium in the Tank Farm, the Effluent Treatment Project, or the Saltstone Production Facility, DWPF has requested an evaluation of the fate of the germanium in the caustic environment of the RCT, the 2H evaporator, and the tank farm. This report evaluates the effect of the addition of germanium to the tank farm based on: (1) the large dilution of Ge in the RCT and tank farm; (2) the solubility of germanium in caustic solutions (pH 12-13); (3) the potential of germanium to precipitate as germanium sodalites in the 2H Evaporator; and (4) the potential of germanium compounds to precipitate in the evaporator feed tank. This study concludes that the impacts of transferring up to 4 kg/yr germanium to the RCT (and subsequently the 2H evaporator feed tank and the 2H evaporator) results in <2 ppm per year (1.834 mg/L) which is the maximum instantaneous concentration expected from DWPF. This concentration is insignificant as most sodium germanates are soluble at the high pH of the feed tank and evaporator solutions. Even if sodium aluminosilicates form in the 2H evaporator, the Ge will likely substitute for some small amount of the Si in these structures and will be insignificant. It is recommended that the DWPF continue with their strategy to add germanium as a laboratory chemical to Attachment 8.2 of the DWPF Waste Compliance Plan (WCP).

Radioactive waste disposal

International Nuclear Information System (INIS)

Bohm, H.; Closs, K.D.; Kuhn, K.

1981-01-01

The solutions to the technical problem of the disposal of radioactive waste are limited by a) the state of knowledge of reprocessing possibilites, b) public acceptance of the use of those techniques which are known, c) legislative procedures linking licensing of new nuclear power plants to the solution of waste problems, and d) other political constraints. Wastes are generated in the mining and enriching of radioactive elements, and in the operation of nuclear power plants as well as in all fields where radioactive substances may be used. Waste management will depend on the stability and concentration of radioactive materials which must be stored, and a resolution of the tension between numerous small storage sites and a few large ones, which again face problems of public acceptability

Guidelines for generators of hazardous chemical waste at LBL and guidelines for generators of radioactive and mixed waste at LBL

International Nuclear Information System (INIS)

1993-10-01

The purpose of this document is to provide the acceptance criteria for the transfer of hazardous chemical waste to LBL's Hazardous Waste Handling Facility (HWHF). Hazardous chemical waste is a necessary byproduct of LBL's research and technical support activities. This waste must be handled properly if LBL is to operate safely and provide adequate protection to staff and the environment. These guidelines describe how you, as a generator of hazardous chemical waste, can meet LBL's acceptance criteria for hazardous chemical waste

Shallow land disposal of radioactive waste

International Nuclear Information System (INIS)

1987-01-01

The application of basic radiation protection concepts and objectives to the disposal of radioactive wastes requires the development of specific reference levels or criteria for the radiological acceptance of each type of waste in each disposal option. This report suggests a methodology for the establishment of acceptance criteria for the disposal of low-level radioactive waste containing long-lived radionuclides in shallow land burial facilities

Alternate paddle configuration for improved wear resistance in the saltstone mixer

Energy Technology Data Exchange (ETDEWEB)

Reigel, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fowley, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2013-09-23

The Saltstone Production Facility has a 10-inch Readco-Kurimoto continuous mixer that mixes the premix dry feeds and low-level waste salt solution to make fresh (uncured) saltstone. Inspection of the mixer in January 2013 showed significant wear on the third, fourth and fifth paddle pairs after the conveying augers. A 2-inch Readco-Kurimoto continuous mixer was used to test alternate paddle configurations for use in the 10-inch mixer to decrease the wear rate on the paddles. Two wear tests were conducted to investigate a method of reducing wear on the mixer paddles. The first test (wear test 2a) had a paddle configuration similar to the currently installed 10-inch mixer in the SPF. This test established baseline wear. The second test (wear test 2b) had a reconfigured paddle arrangement that replaced the flat paddles with helical paddles for paddle pairs 2 - 6 and aligned paddle pair 1 with the augers. The intent of the reconfiguration was to more effectively convey the partially wetted dry feeds through the transition region and into the liquid feed where paddle wear is reduced due to dry feeds and salt solution being mixed at the intended water to premix ratio. The design of the helical paddles provides conveyance through the transition region to the liquid feed inlet. The alignment with the auger is aimed to provide a smoother transition (minimizing the discontinuity between the auger and paddle pair 1) into the downstream paddles. A soft metal with low wear resistance (6000 series aluminum) was used for the wear testing paddles to determine wear patterns while minimizing run time and maximizing wear rate. For the two paddle configurations tested using the scaled 2-inch Readco-Kurimoto continuous mixer, with the first six paddles after the augers replaced by the wear paddles and the remaining paddles were stainless steel. Since the 10-inch SPF mixer is designed with the liquid inlet centered over paddle pairs 5 and 6, the scaled 2-inch mixer was configured the

Salt-stone Oxidation Study: Leaching Method - 13092

International Nuclear Information System (INIS)

Langton, C.A.; Stefanko, D.B.; Burns, H.H.

2013-01-01

Cementitious waste forms can be designed to chemically stabilize selected contaminants, such as Tc +7 and Cr +6 , by chemically reduction to lower valance states, Tc +4 and Cr +3 , respectively, and precipitation of these species in alkaline media as low solubility solid phases. Data for oxidation of this type of cementitious waste form cured under field conditions as a function of time is required for predicting the performance of the waste form and disposal facility. The rate of oxidation (oxidation front advancement) is an important parameter for predicting performance because the solubilities of some radionuclide contaminants, e.g., technetium, are a function of the oxidation state. A non-radioactive experiment was designed for quantifying the oxidation front advancement using chromium, as an approximate redox-sensitive surrogate (Cr +6 / Cr +3 ) for technetium (Tc +7 / Tc +4 ). Nonradioactive cementitious waste forms were prepared in the laboratory and cured under both laboratory and 'field conditions'. Laboratory conditions were ambient temperature and sealed sample containers. Field conditions were approximated by curing samples in open containers which were placed inside a plastic container stored outdoors at SRS. The container had a lid and was instrumented with temperature and humidity probes. Sub-samples as thin as 0.2 mm were taken as a function of distance from the exposed surface of the as-cast sample. The sub-samples were leached and the leachates were analyzed for chromium, nitrate, nitrite and sodium. Nitrate, nitrite, and sodium concentrations were used to provide baseline data because these species are not chemically retained in the waste form matrix to any significant extent and are not redox sensitive. 'Effective' oxidation fronts for Cr were measured for samples containing 1000, 500 and 20 mg/kg Cr added as soluble sodium chromate, Na 2 CrO 4 . For a sample cured for 129 days under field conditions, leachable Cr (assumed to be the oxidized

Acceptance of feed streams for treatment at the LERF/ETF complex

International Nuclear Information System (INIS)

McDonald, F.N.

1994-10-01

This document establishes a waste acceptance process for the storage and treatment of low-level liquid wastes in the LERF and ETF. It also discusses the steps that will be followed when evaluating a waste for storage and treatment

Reconnaissance hydrogeologic investigation of the Defense Waste Processing Facility and Vicinity, Savannah River Plant, South Carolina

International Nuclear Information System (INIS)

Dennehy, K.F.; Prowell, D.C.; McMahon, P.B.

1989-01-01

The purposes of this report are two-fold: (1) to define the hydrogeologic conditions in the vicinity of the defense waste processing facility (DWPF) and, (2) to evaluate the potential for movement of a concentrated salt-solution waste if released at or near the DWPF. These purposes were accomplished by assembling and evaluating existing hydrogeologic data; collecting additional geologic, hydrologic, and water-quality data; developing a local geologic framework; developing a conceptual model of the local ground-water flow system; and by performing laboratory experiments to determine the mobility of salt-solution waste in surface and near-surface sediments. Although the unconsolidated sediments are about 1000 ft thick in the study area, only the Tertiary age sediments, or upper 300 ft are discussed in this report. The top of the Ellenton Formation acts as the major confining unit between the overlying aquifers in Tertiary sediments and the underlying aquifers in Cretaceous sediments; therefore, the Ellenton Formation is the vertical limit of our hydrogeologic investigation. The majority of the hydrologic data for this study come from monitoring wells at the saltstone disposal site (SDS) in Z Area (fig. 3). No recent water-level data were collected in S Area owing to the removal of S Area monitoring wells prior to construction at the DWPF. 46 refs., 26 figs., 7 tabs

BENCH SCALE SALTSTONE PROCESS DEVELOPMENT MIXING STUDY

Energy Technology Data Exchange (ETDEWEB)

Cozzi, A.; Hansen, E.

2011-08-03

The Savannah River National Laboratory (SRNL) was requested to develop a bench scale test facility, using a mixer, transfer pump, and transfer line to determine the impact of conveying the grout through the transfer lines to the vault on grout properties. Bench scale testing focused on the effect the transfer line has on the rheological property of the grout as it was processed through the transfer line. Rheological and other physical properties of grout samples were obtained prior to and after pumping through a transfer line. The Bench Scale Mixing Rig (BSMR) consisted of two mixing tanks, grout feed tank, transfer pump and transfer hose. The mixing tanks were used to batch the grout which was then transferred into the grout feed tank. The contents of the feed tank were then pumped through the transfer line (hose) using a progressive cavity pump. The grout flow rate and pump discharge pressure were monitored. Four sampling stations were located along the length of the transfer line at the 5, 105 and 205 feet past the transfer pump and at 305 feet, the discharge of the hose. Scaling between the full scale piping at Saltstone to bench scale testing at SRNL was performed by maintaining the same shear rate and total shear at the wall of the transfer line. The results of scaling down resulted in a shorter transfer line, a lower average velocity, the same transfer time and similar pressure drops. The condition of flow in the bench scale transfer line is laminar. The flow in the full scale pipe is in the transition region, but is more laminar than turbulent. The resulting plug in laminar flow in the bench scale results in a region of no-mixing. Hence mixing, or shearing, at the bench scale should be less than that observed in the full scale, where this plug is non existent due to the turbulent flow. The bench scale tests should be considered to be conservative due to the highly laminar condition of flow that exists. Two BSMR runs were performed. In both cases, wall

Solvent Carryover Characterization and Recovery for a 10-inch Single Stage Centrifugal Contactor

International Nuclear Information System (INIS)

Lentsch, R.D.; Stephens, A.B.; Leung, D.T.; Baffling, K.E.; Harmon, H.D.; Suggs, P.C.

2006-01-01

A test program has been performed to characterize the organic solvent carryover and recovery from centrifugal contactors in the Caustic-side Solvent Extraction (CSSX) process. CSSX is the baseline design for removing cesium from salt solutions for Department of Energy (DOE) Savannah River Site's Salt Waste Processing Facility. CSSX uses a custom solvent to extract cesium from the salt solution in a series of single stage centrifugal contactors. Meeting the Waste Acceptance Criteria at the Defense Waste Processing Facility and Saltstone, as well as plant economics, dictate that solvent loss should be kept to a minimum. Solvent droplet size distribution in the aqueous outlet streams of the CSSX contactors is of particular importance to the design of solvent recovery equipment. Because insufficient solvent droplet size data existed to form a basis for the recovery system design, DOE funded the CSSX Solvent Carryover Characterization and Recovery Test (SCCRT). This paper presents the droplet size distribution of solvent and concentration in the contactor aqueous outlet streams as a function of rotor speed, bottom plate type, and flow rate. It also presents the performance data of a prototype coalescer. (authors)

Differentiated influences of risk perceptions on nuclear power acceptance according to acceptance targets: Evidence from Korea

International Nuclear Information System (INIS)

Roh, Seung Kook; Lee, Jin Won

2017-01-01

The determinants of the public's nuclear power acceptance have received considerable attention as decisive factors regarding nuclear power policy. However, the contingency of the relative importance of different determinants has been less explored. Building on the literature of psychological distance between the individual and the object, the present study demonstrates that the relative effects of different types of perceived risks regarding nuclear power generation differ across acceptance targets. Using a sample of Korea, our results show that, regarding national acceptance of nuclear power generation, perceived risk from nuclear power plants exerts a stronger negative effect than that from radioactive waste management; however, the latter exerts a stronger negative effect than the former on local acceptance of a nuclear power plant. This finding provides implications for efficient public communication strategy to raise nuclear power acceptance

Assessment of LANL beryllium waste management documentation

International Nuclear Information System (INIS)

Danna, J.G.; Jennrich, E.A.; Lund, D.M.; Davis, K.D.; Hoevemeyer, S.S.

1991-04-01

The objective of this report is to determine present status of the preparation and implementation of the various high priority documents required to properly manage the beryllium waste generated at the Laboratory. The documents being assessed are: Waste Acceptance Criteria, Waste Characterization Plan, Waste Certification Plan, Waste Acceptance Procedures, Waste Characterization Procedures, Waste Certification Procedures, Waste Training Procedures and Waste Recordkeeping Procedures. Beryllium is regulated (as a dust) under 40 CFR 261.33 as ''Discarded commercial chemical products, off specification species, container residues and spill residues thereof.'' Beryllium is also identified in the 3rd thirds ruling of June 1, 1990 as being restricted from land disposal (as a dust). The beryllium waste generated at the Laboratory is handled separately because beryllium has been identified as a highly toxic carcinogenic material

Performance Demonstration Program Plan for Nondestructive Assay of Boxed Wastes for the TRU Waste Characterization Program

International Nuclear Information System (INIS)

2001-01-01

The Performance Demonstration Program (PDP) for nondestructive assay (NDA) consists of a series of tests to evaluate the capability for NDA of transuranic (TRU) waste throughout the Department of Energy (DOE) complex. Each test is termed a PDP cycle. These evaluation cycles provide an objective measure of the reliability of measurements obtained from NDA systems used to characterize the radiological constituents of TRU waste. The primary documents governing the conduct of the PDP are the Waste Acceptance Criteria for the Waste Isolation Pilot Plant (WAC; DOE 1999a) and the Quality Assurance Program Document (QAPD; DOE 1999b). The WAC requires participation in the PDP; the PDP must comply with the QAPD and the WAC. The WAC contains technical and quality requirements for acceptable NDA. This plan implements the general requirements of the QAPD and applicable requirements of the WAC for the NDA PDP for boxed waste assay systems. Measurement facilities demonstrate acceptable performance by the successful testing of simulated waste containers according to the criteria set by this PDP Plan. Comparison among DOE measurement groups and commercial assay services is achieved by comparing the results of measurements on similar simulated waste containers reported by the different measurement facilities. These tests are used as an independent means to assess the performance of measurement groups regarding compliance with established quality assurance objectives (QAO's). Measurement facilities must analyze the simulated waste containers using the same procedures used for normal waste characterization activities. For the boxed waste PDP, a simulated waste container consists of a modified standard waste box (SWB) emplaced with radioactive standards and fabricated matrix inserts. An SWB is a waste box with ends designed specifically to fit the TRUPACT-II shipping container. SWB's will be used to package a substantial volume of the TRU waste for disposal. These PDP sample components

Future-proof radioactive waste treatment technologies for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Buettner, Klaus; Braehler, Georg [NUKEM Technologies Engineering Services GmbH, Alzenau (Germany)

2014-08-15

In order to select the optimal treatment method for radioactive waste three options can be considered. First, to treat the radioactive waste only to allow long term interim storage until the waste acceptance criteria are defined and the disposal sites are operable. Second, to select treatment methods just in compliance with the current state of discussion with the regard to the above. Third, taking also the future development in the field of waste acceptance criteria and disposal into account. When developing waste treatment systems for Nuclear Power Plants NUKEM Technologies follows the following targets, minimisation of the amount of radioactive waste, maximisation of free release material, volume reduction, avoidance of unwanted materials in the waste package, as well as efficient waste treatment solutions (low investment, high volume reduction). With its technologies produced waste packages fulfil the most stringent waste acceptance criteria.

Resource conversation and recovery act draft hazardous waste facility permit: Waste Isolation Pilot Plant (WIPP)

International Nuclear Information System (INIS)

1993-08-01

Volume II contains attachments for Module II and Module III. Attachments for Module II are: part A permit application; examples of acceptable documentation; Waste Isolation Pilot Plant generator/storage site waste screening and acceptance audit program; inspection schedule and monitoring schedule; inspection log forms; personnel training course outlines; hazardous waste job position training requirements; contingency plan; closure plan; and procedures for establishing background for the underground units. One attachment, facility process information, is included for Module III. Remaining attachments for this module are in Volume III

Salt-stone Oxidation Study: Leaching Method - 13092

Energy Technology Data Exchange (ETDEWEB)

Langton, C.A.; Stefanko, D.B.; Burns, H.H. [Savannah River National Laboratory, Savannah River Remediation, LLC, Savannah River Site, Aiken, SC 29808 (United States)

2013-07-01

Cementitious waste forms can be designed to chemically stabilize selected contaminants, such as Tc{sup +7} and Cr{sup +6}, by chemically reduction to lower valance states, Tc{sup +4} and Cr{sup +3}, respectively, and precipitation of these species in alkaline media as low solubility solid phases. Data for oxidation of this type of cementitious waste form cured under field conditions as a function of time is required for predicting the performance of the waste form and disposal facility. The rate of oxidation (oxidation front advancement) is an important parameter for predicting performance because the solubilities of some radionuclide contaminants, e.g., technetium, are a function of the oxidation state. A non-radioactive experiment was designed for quantifying the oxidation front advancement using chromium, as an approximate redox-sensitive surrogate (Cr{sup +6} / Cr{sup +3}) for technetium (Tc{sup +7} / Tc{sup +4}). Nonradioactive cementitious waste forms were prepared in the laboratory and cured under both laboratory and 'field conditions'. Laboratory conditions were ambient temperature and sealed sample containers. Field conditions were approximated by curing samples in open containers which were placed inside a plastic container stored outdoors at SRS. The container had a lid and was instrumented with temperature and humidity probes. Sub-samples as thin as 0.2 mm were taken as a function of distance from the exposed surface of the as-cast sample. The sub-samples were leached and the leachates were analyzed for chromium, nitrate, nitrite and sodium. Nitrate, nitrite, and sodium concentrations were used to provide baseline data because these species are not chemically retained in the waste form matrix to any significant extent and are not redox sensitive. 'Effective' oxidation fronts for Cr were measured for samples containing 1000, 500 and 20 mg/kg Cr added as soluble sodium chromate, Na{sub 2}CrO{sub 4}. For a sample cured for 129 days

Performance Demonstration Program Plan for Nondestructive Assay of Drummed Wastes for the TRU Waste Characterization Program

International Nuclear Information System (INIS)

DOE Carlsbad Field Office

2001-01-01

The Performance Demonstration Program (PDP) for nondestructive assay (NDA) consists of a series of tests to evaluate the capability for NDA of transuranic (TRU) waste throughout the Department of Energy (DOE) complex. Each test is termed a PDP cycle. These evaluation cycles provide an objective measure of the reliability of measurements obtained from NDA systems used to characterize the radiological constituents of TRU waste. The primary documents governing the conduct of the PDP are the Waste Acceptance Criteria for the Waste Isolation Pilot Plant (WAC; DOE 1999a) and the Quality Assurance Program Document (QAPD; DOE 1999b). The WAC requires participation in the PDP; the PDP must comply with the QAPD and the WAC. The WAC contains technical and quality requirements for acceptable NDA. This plan implements the general requirements of the QAPD and applicable requirements of the WAC for the NDA PDP. Measurement facilities demonstrate acceptable performance by the successful testing of simulated waste containers according to the criteria set by this PDP Plan. Comparison among DOE measurement groups and commercial assay services is achieved by comparing the results of measurements on similar simulated waste containers reported by the different measurement facilities. These tests are used as an independent means to assess the performance of measurement groups regarding compliance with established quality assurance objectives (QAO's). Measurement facilities must analyze the simulated waste containers using the same procedures used for normal waste characterization activities. For the drummed waste PDP, a simulated waste container consists of a 55-gallon matrix drum emplaced with radioactive standards and fabricated matrix inserts. These PDP sample components are distributed to the participating measurement facilities that have been designated and authorized by the Carlsbad Field Office (CBFO). The NDA Drum PDP materials are stored at these sites under secure conditions to

Guidelines for generators of hazardous chemical waste at LBL and guidelines for generators of radioactive and mixed waste at LBL

International Nuclear Information System (INIS)

1991-09-01

In part one of this document the Governing Documents and Definitions sections provide general guidelines and regulations applying to the handling of hazardous chemical wastes. The remaining sections provide details on how you can prepare your waste properly for transport and disposal. They are correlated with the steps you must take to properly prepare your waste for pickup. The purpose of the second part of this document is to provide the acceptance criteria for the transfer of radioactive and mixed waste to LBL's Hazardous Waste Handling Facility (HWHF). These guidelines describe how you, as a generator of radioactive or mixed waste, can meet LBL's acceptance criteria for radioactive and mixed waste

Assessment of LANL PCB waste management documentation

International Nuclear Information System (INIS)

David, K.D.; Hoevemeyer, S.S.; Stirrup, T.S.; Jennrich, E.A.; Lund, D.M.

1991-04-01

The objective of this report is to present findings from evaluating the Los Alamos National Laboratory (LANL) Polychlorinated Biphenyls (PCB) Waste Acceptance Criteria (WAC) to determine if it meets applicable DOE and Code of Federal Regulation (CFR) requirements. DOE Order 5820.2A and 40 CFR 761 (Polychlorinated Biphenyls Manufacturing, Processing, Distribution in Commerce, and Use Prohibitions) set forth requirements and guidelines for the establishment of Waste Acceptance Criteria. The primary purpose of a PCB WAC is to provide generators and waste management with established criteria that must be met before PCB wastes can be accepted for treatment, storage, and/or disposal. An annotated outline for a generic PCB WAC was developed based on the requirements of 5820.2A and 40 CFR 761. The major elements that should be addressed by a PCB WAC were determined to be as follows: Waste Package/Container, Waste Forms, PCB Concentrations, Labeling, and Data Package Certification

Packaged low-level waste verification system

Energy Technology Data Exchange (ETDEWEB)

Tuite, K.; Winberg, M.R.; McIsaac, C.V. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

1995-12-31

The Department of Energy through the National Low-Level Waste Management Program and WMG Inc. have entered into a joint development effort to design, build, and demonstrate the Packaged Low-Level Waste Verification System. Currently, states and low-level radioactive waste disposal site operators have no method to independently verify the radionuclide content of packaged low-level waste that arrives at disposal sites for disposition. At this time, the disposal site relies on the low-level waste generator shipping manifests and accompanying records to ensure that low-level waste received meets the site`s waste acceptance criteria. The subject invention provides the equipment, software, and methods to enable the independent verification of low-level waste shipping records to ensure that the site`s waste acceptance criteria are being met. The objective of the prototype system is to demonstrate a mobile system capable of independently verifying the content of packaged low-level waste.

Differentiated influences of risk perceptions on nuclear power acceptance according to acceptance targets: Evidence from Korea

Directory of Open Access Journals (Sweden)

Seungkook Roh

2017-08-01

Full Text Available The determinants of the public's nuclear power acceptance have received considerable attention as decisive factors regarding nuclear power policy. However, the contingency of the relative importance of different determinants has been less explored. Building on the literature of psychological distance between the individual and the object, the present study demonstrates that the relative effects of different types of perceived risks regarding nuclear power generation differ across acceptance targets. Using a sample of Korea, our results show that, regarding national acceptance of nuclear power generation, perceived risk from nuclear power plants exerts a stronger negative effect than that from radioactive waste management; however, the latter exerts a stronger negative effect than the former on local acceptance of a nuclear power plant. This finding provides implications for efficient public communication strategy to raise nuclear power acceptance.

Differentiated influences of risk perceptions on nuclear power acceptance according to acceptance targets: Evidence from Korea

Energy Technology Data Exchange (ETDEWEB)

Roh, Seung Kook [Policy Research Center, Korea Atomic Energy Research Institute (KAERI), Daejeon (Korea, Republic of); Lee, Jin Won [School of Management, Xiamen University, Xiamen (China)

2017-08-15

The determinants of the public's nuclear power acceptance have received considerable attention as decisive factors regarding nuclear power policy. However, the contingency of the relative importance of different determinants has been less explored. Building on the literature of psychological distance between the individual and the object, the present study demonstrates that the relative effects of different types of perceived risks regarding nuclear power generation differ across acceptance targets. Using a sample of Korea, our results show that, regarding national acceptance of nuclear power generation, perceived risk from nuclear power plants exerts a stronger negative effect than that from radioactive waste management; however, the latter exerts a stronger negative effect than the former on local acceptance of a nuclear power plant. This finding provides implications for efficient public communication strategy to raise nuclear power acceptance.

Hanford site transuranic waste certification plan

International Nuclear Information System (INIS)

GREAGER, T.M.

1999-01-01

As a generator of transuranic (TRU) and TRU mixed waste destined for disposal at the Waste Isolation Pilot Plant (WIPP), the Hanford Site must ensure that its TRU waste meets the requirements of U.S. Department of Energy (DOE) Order 5820.2A, ''Radioactive Waste Management, and the Waste Acceptance Criteria for the Waste Isolation Pilot Plant' (DOE 1996d) (WIPP WAC). The WIPP WAC establishes the specific physical, chemical, radiological, and packaging criteria for acceptance of defense TRU waste shipments at WIPP. The WIPP WAC also requires that participating DOE TRU waste generator/treatment/storage sites produce site-specific documents, including a certification plan, that describe their management of TRU waste and TRU waste shipments before transferring waste to WIPP. The Hanford Site must also ensure that its TRU waste destined for disposal at WIPP meets requirements for transport in the Transuranic Package Transporter41 (TRUPACT-11). The U.S. Nuclear Regulatory Commission (NRC) establishes the TRUPACT-I1 requirements in the ''Safety Analysis Report for the TRUPACT-II Shipping Package'' (NRC 1997) (TRUPACT-I1 SARP)

Certification Plan, Radioactive Mixed Waste Hazardous Waste Handling Facility

International Nuclear Information System (INIS)

Albert, R.

1992-01-01

The purpose of this plan is to describe the organization and methodology for the certification of radioactive mixed waste (RMW) handled in the Hazardous Waste Handling Facility at Lawrence Berkeley Laboratory (LBL). RMW is low-level radioactive waste (LLW) or transuranic (TRU) waste that is co-contaminated with dangerous waste as defined in the Westinghouse Hanford Company (WHC) Solid Waste Acceptance Criteria (WAC) and the Washington State Dangerous Waste Regulations, 173-303-040 (18). This waste is to be transferred to the Hanford Site Central Waste Complex and Burial Grounds in Hanford, Washington. This plan incorporates the applicable elements of waste reduction, which include both up-front minimization and end-product treatment to reduce the volume and toxicity of the waste; segregation of the waste as it applies to certification; an executive summary of the Waste Management Quality Assurance Implementing Management Plan (QAIMP) for the HWHF (Section 4); and a list of the current and planned implementing procedures used in waste certification

How is Acceptable Public Risk Determined?

International Nuclear Information System (INIS)

Treichel, Judy

2001-01-01

Acceptance of risk is a value-based decision, that is, the acceptance of risk by a person or group of persons depends on the values of the person or the shared values of the group. In the case of nuclear waste management, the nuclear industry, the regulators, and the general public approach risk from entirely different perspectives, dictated by the separate value systems held by each. The utilities producing radioactive waste view risk assessment as a part of a business decision that involves costs and benefits. The values that drive public acceptance of a national nuclear waste management policy are very different. As stated by Peter Montague of the Environmental Research Foundation: 'The only people I know who are enthusiastic about quantitative risk assessment are people who want to gain permission to expose other humans to dangerous chemicals so someone can make money. Risk assessment has proven to be an effective way to gain the necessary permissions'. Between the industry and the public are the regulators. Most national governments require regulatory agencies to establish rules that provide adequate public safety while allowing industries, whether nuclear or other producers of public commodities, to profitably do business. The general population has always had a fragile relationship with nuclear proponents. There is an atmosphere of mistrust based on the understanding that the values that matter to the general public differ tremendously from those purported by the industry and regulators. The general public is more interested in worst case scenarios; that is, what is the most severe negative consequence to their safety and the safety of their children that could result from nuclear projects. There is no cost or benefit more important to the general public than the health and safety of their families. The rift in values creates a great disparity in proposed solutions to the nuclear waste question. Regulators regard public acceptance of a risk-informed policy

How is Acceptable Public Risk Determined?

Energy Technology Data Exchange (ETDEWEB)

Treichel, Judy [Nevada Nuclear Waste Task Force, Las Vegas, NV (United States)

2001-07-01

Acceptance of risk is a value-based decision, that is, the acceptance of risk by a person or group of persons depends on the values of the person or the shared values of the group. In the case of nuclear waste management, the nuclear industry, the regulators, and the general public approach risk from entirely different perspectives, dictated by the separate value systems held by each. The utilities producing radioactive waste view risk assessment as a part of a business decision that involves costs and benefits. The values that drive public acceptance of a national nuclear waste management policy are very different. As stated by Peter Montague of the Environmental Research Foundation: 'The only people I know who are enthusiastic about quantitative risk assessment are people who want to gain permission to expose other humans to dangerous chemicals so someone can make money. Risk assessment has proven to be an effective way to gain the necessary permissions'. Between the industry and the public are the regulators. Most national governments require regulatory agencies to establish rules that provide adequate public safety while allowing industries, whether nuclear or other producers of public commodities, to profitably do business. The general population has always had a fragile relationship with nuclear proponents. There is an atmosphere of mistrust based on the understanding that the values that matter to the general public differ tremendously from those purported by the industry and regulators. The general public is more interested in worst case scenarios; that is, what is the most severe negative consequence to their safety and the safety of their children that could result from nuclear projects. There is no cost or benefit more important to the general public than the health and safety of their families. The rift in values creates a great disparity in proposed solutions to the nuclear waste question. Regulators regard public acceptance of a risk

Wastes

International Nuclear Information System (INIS)

Bovard, Pierre

The origin of the wastes (power stations, reprocessing, fission products) is determined and the control ensuring the innocuity with respect to man, public acceptance, availability, economics and cost are examined [fr

Radioactive wastes with negligible heat generation suitable for disposal

International Nuclear Information System (INIS)

Brennecke, P.; Schumacher, J.; Warnecke, E.

1987-01-01

It is planned to dispose of radioactive wastes with negligible heat generation in the Konrad repository. Preliminary waste acceptance requirements are derived taking the results of site-specific safety assessments as a basis. These requirements must be fulfilled by the waste packages on delivery. The waste amounts which are currently stored and those anticipated up to the year 2000 are discussed. The disposability of these waste packages in the Konrad repository was evaluated. This examination reveals that basically almost all radioactive wastes with negligible heat generation can be accepted. (orig.) [de

Educational project for the nuclear power and nuclear waste public acceptance fostering

International Nuclear Information System (INIS)

Constantin, M.; Diaconu, Daniela

2005-01-01

Full text: Until now, public acceptance of nuclear power in Romania can be assessed as good. The main motivations could be: the primacy effect of the information, the poverty, the absence of the interest, the relative low informing level. However we expect important changes in the near future: more active NGOs with anti-nuclear opinion, the public's awareness will be more important, serious opposition against the continuation of the nuclear development may appear followed by a reformulation of the Romanian ecologists ideas. The problem of radioactive nuclear wastes (RNW) is the most sensitive and, related to the public opinion, we expect it will become crucial in 10-15 years. The main assumption of our project is that children who are now in schools will be tomorrow's decision makers on 'nuclear energy and RNW. Thus, we intend to prepare the young generation for the future participation in the decision making process related to RNW. The paper shows the main ideas (initial conditions obtained by actual knowledge level measurements, constraints, methods, information contents, evaluation methods and produced outputs) of the educational proposed programme. The work is part of the Romanian contribution to COWAM FP6 project. (authors)

Test Results for CSTR Test 3

International Nuclear Information System (INIS)

Lee, D.D.

2000-01-01

The goal of the Savannah River Salt Waste Processing Program (SPP) is to evaluate and select the most effective technology for the treatment of the high-level waste salt solutions currently being stored in underground storage tanks at the U.S. Department of Energy Savannah River Site (SRS) in Aiken, South Carolina. One of the three technologies currently being developed for this application is the Small-Tank Tetraphenylborate Process (STTP). This process uses sodium tetraphenylborate (NaTPB) to precipitate and remove radioactive cesium from the waste and monosodium titanate (MST) to sorb and remove radioactive strontium and actinides. Oak Ridge National Laboratory is demonstrating this process at the 1:4000 scale using a 20-L capacity continuous-flow stirred-tank reactor (CSTR) system. Since March 1999, three operating campaigns of the 20-L CSTR have been conducted. The ultimate goal is to verify that this process, under certain extremes of operating conditions, can meet the minimum treatment criteria necessary for processing and disposal at the Savannah River Saltstone Facility. The waste acceptance criteria (WAC) for 137 Cs, 90 Sr, and total actinides are 137 Cs and 90 Sr are to obtain decontamination factors (DFs) of 40,000 (99.998% removal) and 26 (96.15% removal), respectively. The DF is defined as the concentration of contaminant in the waste divided by the concentration of contaminant in the effluent stream

Radium bearing waste disposal

International Nuclear Information System (INIS)

Tope, W.G.; Nixon, D.A.; Smith, M.L.; Stone, T.J.; Vogel, R.A.; Schofield, W.D.

1995-01-01

Fernald radium bearing ore residue waste, stored within Silos 1 and 2 (K-65) and Silo 3, will be vitrified for disposal at the Nevada Test Site (NTS). A comprehensive, parametric evaluation of waste form, packaging, and transportation alternatives was completed to identify the most cost-effective approach. The impacts of waste loading, waste form, regulatory requirements, NTS waste acceptance criteria, as-low-as-reasonably-achievable principles, and material handling costs were factored into the recommended approach

The Waste Negotiator's mission

International Nuclear Information System (INIS)

Bataille, Christian

1993-01-01

The mission of the Waste Negotiator is to seek out sites for deep underground laboratories to study their potential for disposal of high level radioactive waste. Although appointed by the government, he acts independently. In 1990, faced by severe public criticism at the way that the waste disposal was being handled, and under increasing pressure to find an acceptable solution, the government stopped the work being carried out by ANDRA (Agence nationale pour la gestion des dechets radioactifs) and initiated a full review of the issues involved. At the same time, parliament also started its own extensive investigation to find a way forward. These efforts finally led to the provision of a detailed framework for the management of long lived radioactive waste, including the construction of two laboratories to investigate possible repository sites. The Waste Negotiator was appointed to carry out a full consultative process in the communities which are considering accepting an underground laboratory. (Author)

Hanford Central Waste Complex: Waste Receiving and Processing Facility dangerous waste permit application

International Nuclear Information System (INIS)

1991-10-01

The Hanford Central Waste Complex is an existing and planned series of treatment, and/or disposal (TSD) unites that will centralize the management of solid waste operations at a single location on the Hanford Facility. The Complex includes two units: the WRAP Facility and the Radioactive Mixed Wastes Storage Facility (RMW Storage Facility). This Part B permit application addresses the WRAP Facility. The Facility will be a treatment and storage unit that will provide the capability to examine, sample, characterize, treat, repackage, store, and certify radioactive and/or mixed waste. Waste treated and stored will include both radioactive and/or mixed waste received from onsite and offsite sources. Certification will be designed to ensure and demonstrate compliance with waste acceptance criteria set forth by onsite disposal units and/or offsite facilities that subsequently are to receive waste from the WRAP Facility. This permit application discusses the following: facility description and general provisions; waste characterization; process information; groundwater monitoring; procedures to prevent hazards; contingency plant; personnel training; exposure information report; waste minimization plan; closure and postclosure requirements; reporting and recordkeeping; other relevant laws; certification

International co-operation in the management of radioactive wastes

Energy Technology Data Exchange (ETDEWEB)

Kang, Changsun [Seoul National University (Korea, Republic of)

1998-07-01

The use of nuclear energy faces with the problem of radioactive waste disposal. The public simply abhors radioactive wastes without rightful cause. Hence, the immediate concern is to overcome this hostile preconception of the public, and to come up with safe and economic ways of disposal which are acceptable to the public. Without it, the public may not accept the idea of further use of nuclear energy. Some have concern about safeguards of fissile plutonium with respect to proliferation while others have concerns regarding protection beyond national borders related to movement of radioactive wastes for foreign disposal. Now, the disposal of radioactive waste is a global problem rather than a problem for individual nation. In this aspect, close international cooperation is recently being brought up more than ever in jointly: improving public acceptance, minimizing waste generation, eliminating the burden on future generations, developing internationally acceptable practices, exploring disposal concepts, and collaborating on R and D. (author)

International co-operation in the management of radioactive wastes

International Nuclear Information System (INIS)

Kang, Changsun

1998-01-01

The use of nuclear energy faces with the problem of radioactive waste disposal. The public simply abhors radioactive wastes without rightful cause. Hence, the immediate concern is to overcome this hostile preconception of the public, and to come up with safe and economic ways of disposal which are acceptable to the public. Without it, the public may not accept the idea of further use of nuclear energy. Some have concern about safeguards of fissile plutonium with respect to proliferation while others have concerns regarding protection beyond national borders related to movement of radioactive wastes for foreign disposal. Now, the disposal of radioactive waste is a global problem rather than a problem for individual nation. In this aspect, close international cooperation is recently being brought up more than ever in jointly: improving public acceptance, minimizing waste generation, eliminating the burden on future generations, developing internationally acceptable practices, exploring disposal concepts, and collaborating on R and D. (author)

Gas characterization system software acceptance test procedure

International Nuclear Information System (INIS)

Vo, C.V.

1996-01-01

This document details the Software Acceptance Testing of gas characterization systems. The gas characterization systems will be used to monitor the vapor spaces of waste tanks known to contain measurable concentrations of flammable gases

Solid, low-level radioactive waste certification program

International Nuclear Information System (INIS)

Grams, W.H.

1991-11-01

The Hanford Site solid waste treatment, storage, and disposal facilities accept solid, low-level radioactive waste from onsite and offsite generators. This manual defines the certification program that is used to provide assurance that the waste meets the Hanford Site waste acceptance criteria. Specifically, this program defines the participation and responsibilities of Westinghouse Hanford Company Solid Waste Engineering Support, Westinghouse Hanford Company Quality Assurance, and both onsite and offsite waste generators. It is intended that waste generators use this document to develop certification plans and quality assurance program plans. This document is also intended for use by Westinghouse Hanford Company solid waste technical staff involved in providing assurance that generators have implemented a waste certification program. This assurance involves review and approval of generator certification plans, and review of generator's quality assurance program plans to ensure that they address all applicable requirements. The document also details the Westinghouse Hanford Company Waste Management Audit and Surveillance Program. 5 refs

Market driven strategy for acquisition of waste acceptance and transportation services for commercial spent fuel in the united states

International Nuclear Information System (INIS)

Lemeshewsky, W.; Macaluso, C.; Smith, P.; Teer, B.

1998-01-01

The Office of Civilian Radioactive Waste Management (OCRWM) in the United States Department of Energy (DOE) has the responsibility under the Nuclear Waste Policy Act of 1982 (the Act) for the shipment of spent nuclear fuel (SNF) from commercial reactors to a Federal facility for storage and/or disposal. The Act requires the use of private industry to the 'fullest extent possible' in the transportation of spent fuels. An OCRWM goal is to develop a safe, efficient and effective transportation system while meeting the mandate of the Act. OCRWM has then develop a strategy for a market driven approach for the acquisition of transportation services and equipment. To implement this strategy, OCRWM is planning to issue a Request for Proposal (RPF) for the provision of the required services and equipment to accept SNF from the utilities and transport the SNF to a Federal facility. Two draft RPFs have been issued with the second draft incorporating comments on the first draft from potential contractors and other interested parties. The overall strategy as outlined in the draft RPF relies on private industry to use the innovative powers of the marketplace to help DOE accomplish its mission objectives. DOE intends to pursue this procurement strategy whether or not the OCRWM program includes interim storage. The concept described in the draft RPF provides for DOE to purchase services and equipment from a contractor-operated waste acceptance and transportation organization. The contractor is expected to provide initial financing for the project including that necessary for initial acquisition of operational equipment, establish the necessary management organization, and mobilize the necessary resources and capabilities to provide the SNF delivery services at a fixed rate. DOE will retain final approval on all routes and maintain primary responsibility to the States, tribes, and local units of government for assuring appropriate interaction and consideration of their input on

Transuranic contaminated waste form characterization and data base

International Nuclear Information System (INIS)

Kniazewycz, B.G.; McArthur, W.C.

1980-07-01

This volume contains 5 appendices. Title listing are: technologies for recovery of transuranics; nondestructive assay of TRU contaminated wastes; miscellaneous waste characteristics; acceptance criteria for TRU waste; and TRU waste treatment technologies

PNGMDR 2013-2015. Management of tritiated wastes from a defective producer

International Nuclear Information System (INIS)

2013-01-01

Tritium-containing wastes require a specific processing. In France, it has been decided to warehouse them during about fifty years to allow a decay of tritium activity of parcels before storing them. After a brief inventory of these wastes, this report addresses wastes which do not comply with the possibilities of a storage centre in terms of acceptance criteria, or compatibility with radiological performance. Thus, the report presents the ANDRA Cires warehousing building, and discusses its limitations regarding the warehousing of tritium-containing wastes. It discusses the conditions of a warehousing of these wastes in the CEA/DAM Valduc centre: criteria for mobilising this warehousing site, description of acceptable wastes, acceptance specification

Toward integrated design of waste management technologies

International Nuclear Information System (INIS)

Carnes, S.A.; Wolfe, A.K.

1994-01-01

Implementation of waste management technologies has been hindered by the intervention of diverse interests. Relying on a perceived history of inadequate and improper management, operations, and technological design, critics have stymied the implementation of scientifically and governmentally approved technologies and facilities, leading to a critical shortage of hazardous, mixed, and radioactive waste management capacity. The research and development (R ampersand D) required to identify technologies that are simultaneously (1) scientifically valid, (2) economically sound, and (3) publicly acceptable must necessarily address, in an integrated and interdisciplinary manner, these three criteria and how best to achieve the integration of stakeholders early in the technology implementation process (i.e., R ampersand D, demonstration, and commercialization). The goal of this paper is to initiate an identification of factors likely to render radioactive and hazardous waste management technologies publicly acceptable and to provide guidance on how technological R ampersand D might be revised to enhance the acceptability of alternative waste management technologies. Principal among these factors are the equitable distribution of costs, risks, and benefits of waste management policies and technologies, the equitable distribution of authority for making waste management policy and selecting technologies for implementation, and the equitable distribution of responsibility for resolving waste management problems. Stakeholder participation in assessing the likely distribution of these factors and mitigative mechanisms to enhance their equitable distribution, together with stakeholder participation in policy and technology R ampersand D, as informed by stakeholder assessments, should enhance the identification of acceptable policies and technologies

Waste generator services implementation plan

Energy Technology Data Exchange (ETDEWEB)

Mousseau, J.; Magleby, M.; Litus, M.

1998-04-01

Recurring waste management noncompliance problems have spurred a fundamental site-wide process revision to characterize and disposition wastes at the Idaho National Engineering and Environmental Laboratory. The reengineered method, termed Waste Generator Services, will streamline the waste acceptance process and provide waste generators comprehensive waste management services through a single, accountable organization to manage and disposition wastes in a timely, cost-effective, and compliant manner. This report outlines the strategy for implementing Waste Generator Services across the INEEL. It documents the culmination of efforts worked by the LMITCO Environmental Management Compliance Reengineering project team since October 1997. These efforts have included defining problems associated with the INEEL waste management process; identifying commercial best management practices; completing a review of DOE Complex-wide waste management training requirements; and involving others through an Integrated Process Team approach to provide recommendations on process flow, funding/charging mechanisms, and WGS organization. The report defines the work that will be performed by Waste Generator Services, the organization and resources, the waste acceptance process flow, the funding approach, methods for measuring performance, and the implementation schedule and approach. Field deployment will occur first at the Idaho Chemical Processing Plant in June 1998. Beginning in Fiscal Year 1999, Waste Generator Services will be deployed at the other major INEEL facilities in a phased approach, with implementation completed by March 1999.

Waste generator services implementation plan

International Nuclear Information System (INIS)

Mousseau, J.; Magleby, M.; Litus, M.

1998-04-01

Recurring waste management noncompliance problems have spurred a fundamental site-wide process revision to characterize and disposition wastes at the Idaho National Engineering and Environmental Laboratory. The reengineered method, termed Waste Generator Services, will streamline the waste acceptance process and provide waste generators comprehensive waste management services through a single, accountable organization to manage and disposition wastes in a timely, cost-effective, and compliant manner. This report outlines the strategy for implementing Waste Generator Services across the INEEL. It documents the culmination of efforts worked by the LMITCO Environmental Management Compliance Reengineering project team since October 1997. These efforts have included defining problems associated with the INEEL waste management process; identifying commercial best management practices; completing a review of DOE Complex-wide waste management training requirements; and involving others through an Integrated Process Team approach to provide recommendations on process flow, funding/charging mechanisms, and WGS organization. The report defines the work that will be performed by Waste Generator Services, the organization and resources, the waste acceptance process flow, the funding approach, methods for measuring performance, and the implementation schedule and approach. Field deployment will occur first at the Idaho Chemical Processing Plant in June 1998. Beginning in Fiscal Year 1999, Waste Generator Services will be deployed at the other major INEEL facilities in a phased approach, with implementation completed by March 1999

Nuclear waste

International Nuclear Information System (INIS)

Anon.

1996-01-01

The NEA Nuclear Waste Bulletin has been prepared by the Radiation Protection and Waste Management Division of the OECD Nuclear Energy Agency to provide a means of communication amongst the various technical and policy groups within the waste management community. In particular, it is intended to provide timely and concise information on radioactive waste management activities, policies and programmes in Member countries and at the NEA. It is also intended that the Bulletin assists in the communication of recent developments in a variety of areas contributing to the development of acceptable technology for the management and disposal of nuclear waste (e.g., performance assessment, in-situ investigations, repository engineering, scientific data bases, regulatory developments, etc)

Conditions governing the acceptance of radioactive wastes by the Hauptabteilung Dekontaminationsbetriebe (HDB). Full text of legal provisions, issue no.6 of July 1, 1991, as amended until January 1, 1995

International Nuclear Information System (INIS)

1995-01-01

The conditions apply to the acceptance of radwaste by the Main Decontamination Dept. (HDB) of Karlsruhe Research Center, including radioactive remnants, contaminated plant components, and primary waste from the following waste generators: Institutes of the Karlsruhe Research Center, facilities located within the Center but run by other organisations, other outside facilities not linked with the Center, as e.g. waste generators in Baden-Wuerttemberg obliged to deliver their radwaste to the Radwaste Collecting Site of the Land of Baden-Wuerttemberg. Amendments are marked at the right-hand margin of the text

Status of waste form testing

International Nuclear Information System (INIS)

Lawroski, H.

1984-01-01

The promulgation of the amendment of 10 CFR Part 61 by the Nuclear Regulatory Commission of December 27, 1982 by Federal Register Notice with an effective date of December 27, 1983 established the criteria for licensing requirements, paragraph 60.56, contained the description to provide adequate stability of the site through the use of suitable waste forms. In May, 1983, the NRC published a final Branch Technical Position (BTP) paper on waste form. The position taken by the BTP was considerably more severe than indicated in 10 CFR Part 61. An extensive and expensive testing program was started in 1983. As an interim measure, the presently utilized solidification processes such as cement, Dow binder, Envirostone and bitumen, and the presently qualified High Integrity containers (HICs) were considered acceptable with the caveat that acceptable process control programs were being utilized. The NRC requested that topical reports for licenses be submitted. The topical reports were to contain test results to substantiate the acceptability of the waste forms. The test results to date show that the volume of wastes will have to increase to meet the position taken by the NRC in the BTP. This position will cause more waste to be generated which is contrary to the emphasis by states and others to reduce the volume of waste. The details of testing will be discussed in the paper to be presented

Gas characterization system software acceptance test report

International Nuclear Information System (INIS)

Vo, C.V.

1996-01-01

This document details the results of software acceptance testing of gas characterization systems. The gas characterization systems will be used to monitor the vapor spaces of waste tanks known to contain measurable concentrations of flammable gases

Solid Waste Burial Grounds/Central Waste Complex hazards assessment

International Nuclear Information System (INIS)

Broz, R.E.

1994-01-01

This document establishes the technical basis in support of Emergency Planning Activities for Solid Waste Burial Grounds/Central Waste Complex on the Hanford Site. The document represents an acceptable interpretation of the implementing guidance document for DOE Order 5500.3A. Through this document, the technical basis for the development of facility specific Emergency Action Levels and the Emergency Planning Zone is documented

The Savannah River Site Waste Inventory Management Program

International Nuclear Information System (INIS)

Griffith, J.M.; Holmes, B.R.

1995-01-01

Each hazardous and radioactive waste generator that delivers waste to Savannah River Site (SRS) treatment, storage and disposal (TSD) facilities is required to implement a waste certification plan. The waste certification process ensures that waste has been properly identified, characterized, segregated, packaged, and shipped according to the receiving facilities waste acceptance criteria. In order to comply with the rigid acceptance criteria, the Reactor Division developed and implemented the Waste Inventory Management Program (WIMP) to track the generation and disposal of low level radioactive waste. The WIMP system is a relational database with integrated barcode technology designed to track the inventory radioactive waste. During the development of the WIMP several waste minimization tools were incorporated into the design of the program. The inclusion of waste minimization tools as part of the WIMP has resulted in a 40% increase in the amount of waste designated as compactible and an overall volume reduction of 5,000 cu-ft

Nuclear waste - perceptions and realities

International Nuclear Information System (INIS)

Wilkinson, D.

1984-01-01

This paper discusses the complex scientific, sociological, political and emotive aspects of nuclear waste. The public perception of the hazards and risks, to present and future generations, in the management of nuclear wastes are highlighted. The cost of nuclear waste management to socially acceptable and technically achievable standards is discussed. (UK)

Overview of the NRC nuclear waste management program

International Nuclear Information System (INIS)

Malaro, J.C.

1976-01-01

The NRC has firmly established waste management as a high-priority effort and has made the commitment to act rapidly and methodically to establish a sound regulatory base for licensing waste management activities. We believe the priorities for NRC work in waste management are consistent with the needs of the overall national waste management program. Present licensing procedures and criteria are adequate for the short term, and priority attention is being given to the longer term, when the quantities of waste to be managed will be greater and licensing demands will increase. Recognizing that its decision will affect industry, other governmental jurisdictions, private interest groups, and the public at large, NRC has encouraged and will continue to encourage their participation in planning our program. We also recognize that the problems of nuclear waste management are international in scope. Many waste management problems (e.g., potential for contamination of oceans and atmosphere, need for isolation of some wastes for longer periods than governments and political boundaries have remained stable in the past), require a set of internationally acceptable and accepted solutions. The wastes from the U.S. nuclear industry will account for only about one third of the nuclear waste generated in the world. Therefore, we propose to cooperate and where appropriate take the lead in establishing acceptable worldwide policies, standards and procedures for handling nuclear wastes

Calculation of projected waste loads for transuranic waste management alternatives

International Nuclear Information System (INIS)

Hong, K.; Kotek, T.; Koebnick, B.; Wang, Y.; Kaicher, C.

1995-01-01

The level of treatment and the treatment and interim storage site configurations (decentralized, regional, or centralized) impact transuranic (TRU) waste loads at and en route to sites in the US Department of Energy (DOE) complex. Other elements that impact waste loads are the volume and characteristics of the waste and the unit operation parameters of the technologies used to treat it. Projected annual complexwide TRU waste loads under various TRU waste management alternatives were calculated using the WASTEunderscoreMGMT computational model. WASTEunderscoreMGMT accepts as input three types of data: (1) the waste stream inventory volume, mass, and contaminant characteristics by generating site and waste stream category; (2) unit operation parameters of treatment technologies; and (3) waste management alternative definitions. Results indicate that the designed capacity of the Waste Isolation Pilot Plant, identified under all waste management alternatives as the permanent disposal facility for DOE-generated TRU waste, is sufficient for the projected complexwide TRU waste load under any of the alternatives

Vitrification of hazardous and radioactive wastes

International Nuclear Information System (INIS)

Bickford, D.F.; Schumacher, R.

1995-01-01

Vitrification offers many attractive waste stabilization options. Versatility of waste compositions, as well as the inherent durability of a glass waste form, have made vitrification the treatment of choice for high-level radioactive wastes. Adapting the technology to other hazardous and radioactive waste streams will provide an environmentally acceptable solution to many of the waste challenges that face the public today. This document reviews various types and technologies involved in vitrification

Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project. Companion CD-ROM

International Nuclear Information System (INIS)

2016-05-01

Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants