Corrosion control for the Hanford site waste transfer system

International Nuclear Information System (INIS)

Haberman, J.H.

1995-01-01

Processing large volumes of spent reactor fuel and other related waste management activities produced radioactive wastes which have been stored in underground high-level waste storage tanks since the 1940s. The effluent waste streams from the processing facilities were stored underground in high-level waste storage tanks. The waste was transferred between storage tanks and from the tanks to waste processing facilities in a complex network of underground piping. The underground waste transfer system consists of process piping, catch tanks, lift tanks, diversion boxes, pump pits, valves, and jumpers. Corrosion of the process piping from contact with the soil is a primary concern. The other transfer system components are made of corrosion-resistant alloys or they are isolated from the underground environment and experience little degradation. Corrosion control of the underground transfer system is necessary to ensure that transfer routes will be available for future waste retrieval, processing,a nd disposal. Today, most waste transfer lines are protected by an active impressed-current cathodic protection (CP) system. The original system has been updated. Energization surveys and a recent base-line survey demonstrate that system operational goals are met

Waste classification sampling plan

International Nuclear Information System (INIS)

Landsman, S.D.

1998-01-01

The purpose of this sampling is to explain the method used to collect and analyze data necessary to verify and/or determine the radionuclide content of the B-Cell decontamination and decommissioning waste stream so that the correct waste classification for the waste stream can be made, and to collect samples for studies of decontamination methods that could be used to remove fixed contamination present on the waste. The scope of this plan is to establish the technical basis for collecting samples and compiling quantitative data on the radioactive constituents present in waste generated during deactivation activities in B-Cell. Sampling and radioisotopic analysis will be performed on the fixed layers of contamination present on structural material and internal surfaces of process piping and tanks. In addition, dose rate measurements on existing waste material will be performed to determine the fraction of dose rate attributable to both removable and fixed contamination. Samples will also be collected to support studies of decontamination methods that are effective in removing the fixed contamination present on the waste. Sampling performed under this plan will meet criteria established in BNF-2596, Data Quality Objectives for the B-Cell Waste Stream Classification Sampling, J. M. Barnett, May 1998

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

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

Soluble pig for radioactive waste transfer lines

International Nuclear Information System (INIS)

Ohl, P.C.; Pezeshki, C.

1997-01-01

Flushing transfer pipe after radioactive waste transfers generates thousands of gallons of additional radioactive waste each year at the Hanford site. The use of pneumatic pigging with waste soluble pigs as a means to clear transfer piping may be an effective alternative to raw water flushes. A feasibility study was performed by a group of senior mechanical engineering students for their senior design project as part of their curriculum at Washington State University. The students divided the feasibility study into three sub-projects involving: (1) material research, (2) delivery system design, and (3) mockup fabrication and testing. The students screened through twenty-three candidate materials and selected a thermoplastic polymer combined 50:50 wt% with sucrose to meet the established material performance criteria. The students also prepared a conceptual design of a remote pneumatic delivery system and constructed a mockup section of transfer pipe for testing the prototype pigs

Liquid waste sampling device

International Nuclear Information System (INIS)

Kosuge, Tadashi

1998-01-01

A liquid pumping pressure regulator is disposed on the midway of a pressure control tube which connects the upper portion of a sampling pot and the upper portion of a liquid waste storage vessel. With such a constitution, when the pressure in the sampling pot is made negative, and liquid wastes are sucked to the liquid pumping tube passing through the sampling pot, the difference between the pressure on the entrance of the liquid pumping pressure regulator of the pressure regulating tube and the pressure at the bottom of the liquid waste storage vessel is made constant. An opening degree controlling meter is disposed to control the degree of opening of a pressure regulating valve for sending actuation pressurized air to the liquid pumping pressure regulator. Accordingly, even if the liquid level of liquid wastes in the liquid waste storage vessel is changed, the height for the suction of the liquid wastes in the liquid pumping tube can be kept constant. With such procedures, sampling can be conducted correctly, and the discharge of the liquid wastes to the outside can be prevented. (T.M.)

Material Balance Assessment for Double-Shell Tank Waste Pipeline Transfer

International Nuclear Information System (INIS)

Onishi, Yasuo; Wells, Beric E; Hartley, Stacey A; Enderlin, Carl W

2001-01-01

PNNL developed a material balance assessment methodology based on conservation of mass for detecting leaks and mis-routings in pipeline transfer of double-shell tank waste at Hanford. The main factors causing uncertainty in these transfers are variable property and tank conditions of density, existence of crust, and surface disturbance due to mixer pump operation during the waste transfer. The methodology was applied to three waste transfers from Tanks AN-105 and AZ-102

Optimization of waste transportation route at waste transfers point in Lowokwaru District, Malang City

Science.gov (United States)

Hariyani, S.; Meidiana, C.

2018-04-01

Increasing population led to the emergence of the urban infrastructure services issue including waste problems especially waste transportation system. Data in 2016 shows that the amount of waste in Malang was 659.21 tons / day. The amount of waste transported to landfill only reached 464.74 tons / day. This indicates that not all waste can be transported to the landfill Supiturang because Level of Service (LoS) reached 70.49%. This study aims to determine the effectiveness of waste transportation system and determine the fastest route from waste transfers point in Lowokwaru district to the landfill Supiturang. The data collection method in this research were 1) primary survey by interview officials from the Sanitation and Gardening Agency which questions related to the condition of the waste transportation system in waste transfer point, 2) Secondary survey related to data of waste transportation system in Malang City i.e the amount of waste generation in waste transfer point, number of garbage trucks and other data related to the garbage transportation system. To determine the fastest route analyzed by network analyst using ArcGIS software. The results of network analyst show that not all routes are already using the fastest route to the landfill Supiturang.

Waste conditioning for tank heel transfer. Preliminary data and results

International Nuclear Information System (INIS)

Ebadian, M.A.

1999-01-01

This report summarizes the research carried out at Florida International University's Hemispheric Center for Environmental Technology (FIU-HCET) for the fiscal year 1998 (FY98) under the Tank Focus Area (TFA) project ''Waste Conditioning for Tank Slurry Transfer.'' The objective of this project is to determine the effect of chemical and physical properties on the waste conditioning process and transfer. The focus of this research consisted in building a waste conditioning experimental facility to test different slurry simulants under different conditions, and analyzing their chemical and physical properties. This investigation would provide experimental data and analysis results that can make the tank waste conditioning process more efficient, improve the transfer system, and influence future modifications to the waste conditioning and transfer system. A waste conditioning experimental facility was built in order to test slurry simulants. The facility consists of a slurry vessel with several accessories for parameter control and sampling. The vessel also has a lid system with a shaft-mounted propeller connected to an air motor. In addition, a circulation system is connected to the slurry vessel for simulant cooling and heating. Experimental data collection and analysis of the chemical and physical properties of the tank slurry simulants has been emphasized. For this, one waste slurry simulant (Fernald) was developed, and another two simulants (SRS and Hanford) obtained from DOE sites were used. These simulants, composed of water, soluble metal salts, and insoluble solid particles, were used to represent the actual radioactive waste slurries from different DOE sites. The simulants' chemical and physical properties analyzed include density, viscosity, pH, settling rate, and volubility. These analyses were done to samples obtained from different experiments performed at room temperature but different mixing time and strength. The experimental results indicate that the

Small Scale Mixing Demonstration Batch Transfer and Sampling Performance of Simulated HLW - 12307

Energy Technology Data Exchange (ETDEWEB)

Jensen, Jesse; Townson, Paul; Vanatta, Matt [EnergySolutions, Engineering and Technology Group, Richland, WA, 99354 (United States)

2012-07-01

The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste treatment Plant (WTP) has been recognized as a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. At the end of 2009 DOE's Tank Operations Contractor, Washington River Protection Solutions (WRPS), awarded a contract to EnergySolutions to design, fabricate and operate a demonstration platform called the Small Scale Mixing Demonstration (SSMD) to establish pre-transfer sampling capacity, and batch transfer performance data at two different scales. This data will be used to examine the baseline capacity for a tank mixed via rotational jet mixers to transfer consistent or bounding batches, and provide scale up information to predict full scale operational performance. This information will then in turn be used to define the baseline capacity of such a system to transfer and sample batches sent to WTP. The Small Scale Mixing Demonstration (SSMD) platform consists of 43'' and 120'' diameter clear acrylic test vessels, each equipped with two scaled jet mixer pump assemblies, and all supporting vessels, controls, services, and simulant make up facilities. All tank internals have been modeled including the air lift circulators (ALCs), the steam heating coil, and the radius between the wall and floor. The test vessels are set up to simulate the transfer of HLW out of a mixed tank, and collect a pre-transfer sample in a manner similar to the proposed baseline configuration. The collected material is submitted to an NQA-1 laboratory for chemical analysis. Previous work has been done to assess tank mixing performance at both scales. This work involved a combination of unique instruments to understand the three dimensional distribution of solids using a combination of Coriolis meter measurements, in situ chord length distribution

Technology transfer and the management of radioactive waste

International Nuclear Information System (INIS)

Bonne, A.; Chan-Sands, C.

1998-01-01

One of the IAEA's fundamental roles is to act as a centre for the transfer of nuclear technologies, including those for managing radioactive wastes. In the area of waste management technology, the Agency is actively working to improve and develop new and efficient means to fulfill that responsibility. Recognizing its responsibilities and challenges, IAEA efforts related to radioactive waste management technologies into the next century are framed around three major areas: the development and implementation of mechanisms for better technology transfer and information exchange; the promotion of sustainable and safer processes and procedures; and the provision of peer reviews and direct technical assistance that help facilitate bilateral and multinational efforts. To illustrate some specific elements of the overall programme, this article reviews selected technology-transfer activities that have been initiated in the field

Evaluation of tank waste transfers at 241-AW tank farm

International Nuclear Information System (INIS)

Willis, W.L.

1998-01-01

A number of waste transfers are needed to process and feed waste to the private contractors in support of Phase 1 Privatization. Other waste transfers are needed to support the 242-A Evaporator, saltwell pumping, and other ongoing Tank Waste Remediation System (TWRS) operations. The purpose of this evaluation is to determine if existing or planned equipment and systems are capable of supporting the Privatization Mission of the Tank Farms and continuing operations through the end of Phase 1B Privatization Mission. Projects W-211 and W-314 have been established and will support the privatization effort. Equipment and system upgrades provided by these projects (W-211 and W-314) will also support other ongoing operations in the tank farms. It is recognized that these projects do not support the entire transfer schedule represented in the Tank Waste Remediation system Operation and Utilization Plan. Additionally, transfers surrounding the 241-AW farm must be considered. This evaluation is provided as information, which will help to define transfer paths required to complete the Waste Feed Delivery (WFD) mission. This document is not focused on changing a particular project, but it is realized that new project work in the 241-AW Tank Farm is required

Initial Investigation of Waste Feed Delivery Tank Mixing and Sampling Issues

International Nuclear Information System (INIS)

Fort, James A.; Bamberger, Judith A.; Meyer, Perry A.; Stewart, Charles W.

2007-01-01

The Hanford tank farms contractor will deliver waste to the Waste Treatment Plant (WTP) from a staging double-shell tank. The WTP broadly classifies waste it receives in terms of 'Envelopes,' each with different limiting properties and composition ranges. Envelope A, B, and C wastes are liquids that can include up to 4% entrained solids that can be pumped directly from the staging DST without mixing. Envelope D waste contains insoluble solids and must be mixed before transfer. The mixing and sampling issues lie within Envelope D solid-liquid slurries. The question is how effectively these slurries are mixed and how representative the grab samples are that are taken immediately after mixing. This report summarizes the current state of knowledge concerning jet mixing of wastes in underground storage tanks. Waste feed sampling requirements are listed, and their apparent assumption of uniformity by lack of a requirement for sample representativeness is cited as a significant issue. The case is made that there is not an adequate technical basis to provide such a sampling regimen because not enough is known about what can be achieved in mixing and distribution of solids by use of the baseline submersible mixing pump system. A combined mixing-sampling test program is recommended to fill this gap. Historical Pacific Northwest National Laboratory project and tank farms contractor documents are used to make this case. A substantial investment and progress are being made to understand mixing issues at the WTP. A summary of the key WTP activities relevant to this project is presented in this report. The relevant aspects of the WTP mixing work, together with a previously developed scaled test strategy for determining solids suspension with submerged mixer pumps (discussed in Section 3) provide a solid foundation for developing a path forward

Development of bagless transfer system for standard waste boxes

International Nuclear Information System (INIS)

Presgrove, S.B.; Patel, K.

1991-01-01

At several DOE site, substantial volumes of Transuranic (TRU) waste has been handled on a daily basis. Usually, the waste has been transferred from the facilities to the Waste Isolation Pilot Plant (WIPP) or to an on site storage facility using the ''bag-out'' technique. This process begins in the most contaminated area by placing the waste in a strong plastic bag, twisting closed the neck of the bag, then taping the closed neck. This reduces the exposed TRU waste and the resulting contamination. However, even though that the contamination is reduced, it remains high enough to prevent direct transfer into the environment. In order to reduce the contamination to acceptable levels, the ''bag-out'' process is repeated until the outside contamination on the plastic bag is acceptable. This procedure has been affective, however, it also generates organic waste in the process. During the design of the Transuranic Waste Facility (TWF) at the Savannah River Site, a conceptual alternative was developed using the Standard Waste Box (SWB)

Control Decisions for Flammable Gas Hazards in Waste Transfer Systems

International Nuclear Information System (INIS)

KRIPPS, L.J.

2000-01-01

This report describes the control decisions for flammable gas hazards in waste transfer systems (i.e., waste transfer piping and waste transfer-associated structures) made at control decision meetings on November 30, 1999a and April 19, 2000, and their basis. These control decisions, and the analyses that support them, will be documented in an amendment to the Final Safety Analysis Report (FSAR) (CHG 2000a) and Technical Safety Requirements (TSR) (CHG 2000b) to close the Flammable Gas Unreviewed Safety Question (USQ) (Bacon 1996 and Wagoner 1996). Following the Contractor Tier I review of the FSAR and TSR amendment, it will be submitted to the US. Department of Energy (DOE), Office of River Protection (ORP) for review and approval. The control decision meeting on November 30, 1999 to address flammable gas hazards in waste transfer systems followed the control decision process and the criteria for control decisions described in Section 3.3.1.5 of the FSAR. The control decision meeting agenda, attendance list, and introductory and background presentations are included in Attachments 1 through 4. The control decision discussions on existing and other possible controls for flammable gas hazards in waste transfer systems and the basis for selecting or not selecting specific controls are summarized in this report

TRU waste-sampling program

International Nuclear Information System (INIS)

Warren, J.L.; Zerwekh, A.

1985-08-01

As part of a TRU waste-sampling program, Los Alamos National Laboratory retrieved and examined 44 drums of 238 Pu- and 239 Pu-contaminated waste. The drums ranged in age from 8 months to 9 years. The majority of drums were tested for pressure, and gas samples withdrawn from the drums were analyzed by a mass spectrometer. Real-time radiography and visual examination were used to determine both void volumes and waste content. Drum walls were measured for deterioration, and selected drum contents were reassayed for comparison with original assays and WIPP criteria. Each drum tested at atmospheric pressure. Mass spectrometry revealed no problem with 239 Pu-contaminated waste, but three 8-month-old drums of 238 Pu-contaminated waste contained a potentially hazardous gas mixture. Void volumes fell within the 81 to 97% range. Measurements of drum walls showed no significant corrosion or deterioration. All reassayed contents were within WIPP waste acceptance criteria. Five of the drums opened and examined (15%) could not be certified as packaged. Three contained free liquids, one had corrosive materials, and one had too much unstabilized particulate. Eleven drums had the wrong (or not the most appropriate) waste code. In many cases, disposal volumes had been inefficiently used. 2 refs., 23 figs., 7 tabs

Hanford site transuranic waste sampling plan

International Nuclear Information System (INIS)

GREAGER, T.M.

1999-01-01

This sampling plan (SP) describes the selection of containers for sampling of homogeneous solids and soil/gravel and for visual examination of transuranic and mixed transuranic (collectively referred to as TRU) waste generated at the U.S. Department of Energy (DOE) Hanford Site. The activities described in this SP will be conducted under the Hanford Site TRU Waste Certification Program. This SP is designed to meet the requirements of the Transuranic Waste Characterization Quality Assurance Program Plan (CAO-94-1010) (DOE 1996a) (QAPP), site-specific implementation of which is described in the Hanford Site Transuranic Waste Characterization Program Quality Assurance Project Plan (HNF-2599) (Hanford 1998b) (QAPP). The QAPP defines the quality assurance (QA) requirements and protocols for TRU waste characterization activities at the Hanford Site. In addition, the QAPP identifies responsible organizations, describes required program activities, outlines sampling and analysis strategies, and identifies procedures for characterization activities. The QAPP identifies specific requirements for TRU waste sampling plans. Table 1-1 presents these requirements and indicates sections in this SP where these requirements are addressed

Verification of Representative Sampling in RI waste

International Nuclear Information System (INIS)

Ahn, Hong Joo; Song, Byung Cheul; Sohn, Se Cheul; Song, Kyu Seok; Jee, Kwang Yong; Choi, Kwang Seop

2009-01-01

For evaluating the radionuclide inventories for RI wastes, representative sampling is one of the most important parts in the process of radiochemical assay. Sampling to characterized RI waste conditions typically has been based on judgment or convenience sampling of individual or groups. However, it is difficult to get a sample representatively among the numerous drums. In addition, RI waste drums might be classified into heterogeneous wastes because they have a content of cotton, glass, vinyl, gloves, etc. In order to get the representative samples, the sample to be analyzed must be collected from selected every drum. Considering the expense and time of analysis, however, the number of sample has to be minimized. In this study, RI waste drums were classified by the various conditions of the half-life, surface dose, acceptance date, waste form, generator, etc. A sample for radiochemical assay was obtained through mixing samples of each drum. The sample has to be prepared for radiochemical assay and although the sample should be reasonably uniform, it is rare that a completely homogeneous material is received. Every sample is shredded by a 1 ∼ 2 cm 2 diameter and a representative aliquot taken for the required analysis. For verification of representative sampling, classified every group is tested for evaluation of 'selection of representative drum in a group' and 'representative sampling in a drum'

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

Radiological safety evaluation for a Waste Transfer Facility at Savannah River Site

International Nuclear Information System (INIS)

Ades, M.J.

1993-01-01

This paper provides a review of the radiological safety evaluation performed for a Waste Transfer Facility (WTF) located at the Savannah River Site (SRS). This facility transfers liquid radioactive waste between various waste processing facilities and waste storage facilities. The WTF includes functional components such as the diversion box and the pump pits, waste transfer lines, and the outside yard service piping and electrical services. The WSRC methodology is used to evaluate the consequences of postulated accidents that result in the release of radioactive material. Such accidents include transfer line breaks, underground liquid pathway release, fire in pump tank cells and HEPA filters, accidents due to natural phenomena, and externally induced events. Chemical hazards accidents are not considered. The analysis results indicate that the calculated mean onsite and offsite radiological consequences are bounded by the corresponding WSRC dose limits for each accident considered. Moreover, the results show that the maximum onsite and offsite doses calculated for the WTF are lower than the maximum doses determined for the whole radioactive waste facility where the WTF is located

New Equipment and Techniques for Remote Sampling of Stored Radioactive Waste

International Nuclear Information System (INIS)

Nance, T.A.

2001-01-01

Radioactive waste is stored at the Savannah River Site (SRS), part of the Department of Energy (DOE) complex. This radioactive waste is stored in buried tanks and management of the waste requires several processes, including material addition, heating, cooling, mixing, and transfer from tank to tank. During waste processing, it is necessary to know the chemical components and their characteristics to determine the steps necessary to maintain the waste form or to manipulate the waste into the form desired. Waste characterization begins by obtaining a sample for analysis. High level radioactive waste sampling is routinely done with simple, standard samplers such as a dip sample. Other sampling is non-routine or specialized, with unique, special requirements, such as sampling remote areas that are difficult to reach. Other specialized sampling includes sampling materials with unknown characteristics or material that must be gathered to obtain an adequate sample or materials that must be broken up to sample or forcibly separated from the tank. The samplers described in this paper are specialized samplers. These samplers include the Dip Filter Sampler, Soft Core Sampler, Hard Core Sampler, Circle Scrape Sampler, Small Scrape Sampler, Suction and Strain Sampler, and Vial Snapper Sampler. The Dip Filter Sampler is used to sample floating particulate matter or floating organic matter. The Soft Core Sampler and Hard Core Sampler are used to obtain samples of solids from the tank floor. The Soft Core Sampler is used on soft solids such as sludge and saltcake and the Hard Core Sampler on hardened solid deposits. The Circle Scrape Sampler is used to obtain solid samples through a small entry riser and out from under the riser. The Small Scrape Sampler enters a small entry riser and is used to scrape a sample from the tank wall. The Suction and Strain Sampler is used to gather a remote submerged sample or filter a solid sample from supernate. The Vial Snapper Grab Sampler is

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

A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests

International Nuclear Information System (INIS)

Thien, Mike G.; Barnes, Steve M.

2013-01-01

The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described

Railcar waste transfer system hydrostatic test

International Nuclear Information System (INIS)

Ellingson, S.D.

1997-01-01

Recent modifications have been performed on the T-Plant Railcar Waste Transfer System, This Acceptance Test Procedure (ATP) has been prepared to demonstrate that identified piping welds and mechanical connections incorporated during the modification are of high integrity and are acceptable for service. This will be achieved by implementation of a hydrostatic leak test

Polybrominated diphenyl ethers in e-waste: Level and transfer in a typical e-waste recycling site in Shanghai, Eastern China

Energy Technology Data Exchange (ETDEWEB)

Li, Yue; Duan, Yan-Ping, E-mail: duanyanping@tongji.edu.cn; Huang, Fan; Yang, Jing; Xiang, Nan; Meng, Xiang-Zhou; Chen, Ling

2014-06-01

Highlights: • PBDEs were detected in the majority of e-waste. • PBDEs were found in TVs made in China after 1990. • The levels of ΣPBDEs in e-waste made in Japan far exceed the threshold limit of RoHS. • The inappropriate recycling and disposal of e-waste is an important source of PBDEs. - Abstract: Very few data for polybrominated diphenyl ethers (PBDEs) were available in the electronic waste (e-waste) as one of the most PBDEs emission source. This study reported concentrations of PBDEs in e-waste including printer, rice cooker, computer monitor, TV, electric iron and water dispenser, as well as dust from e-waste, e-waste dismantling workshop and surface soil from inside and outside of an e-waste recycling plant in Shanghai, Eastern China. The results showed that PBDEs were detected in the majority of e-waste, and the concentrations of ΣPBDEs ranged from not detected to 175 g/kg, with a mean value of 10.8 g/kg. PBDEs were found in TVs made in China after 1990. The mean concentrations of ΣPBDEs in e-waste made in Korea, Japan, Singapore and China were 1.84 g/kg, 20.5 g/kg, 0.91 g/kg, 4.48 g/kg, respectively. The levels of ΣPBDEs in e-waste made in Japan far exceed the threshold limit of RoHS (1.00 g/kg). BDE-209 dominated in e-waste, accounting for over 93%. The compositional patterns of PBDEs congeners resembled the profile of Saytex 102E, indicating the source of deca-BDE. Among the samples of dust and surface soil from a typical e-waste recycling site, the highest concentrations of Σ{sub 18}PBDEs and BDE-209 were found in dust in e-waste, ranging from 1960 to 340,710 ng/g and from 910 to 320,400 ng/g, which were 1–2 orders of magnitude higher than other samples. It suggested that PBDEs released from e-waste via dust, and then transferred to surrounding environment.

Sampling and characterization of radioactive liquid wastes

International Nuclear Information System (INIS)

Zepeda R, C.; Monroy G, F.; Reyes A, T.; Lizcano, D.; Cruz C, A. C.

2017-09-01

To define the management of radioactive liquid wastes stored in 200 L drums, its isotope and physicochemical characterization is essential. An adequate sampling, that is, representative and homogeneous, is fundamental to obtain reliable analytical results, therefore, in this work, the use of a sampling mechanism that allows collecting homogenous aliquots, in a safe way and minimizing the generation of secondary waste is proposed. With this mechanism, 56 drums of radioactive liquid wastes were sampled, which were characterized by gamma spectrometry, liquid scintillation, and determined the following physicochemical properties: ph, conductivity, viscosity, density and chemical composition by gas chromatography. 67.86% of the radioactive liquid wastes contains H-3 and of these, 47.36% can be released unconditionally, since it presents activities lower than 100 Bq/g. 94% of the wastes are acidic and 48% have viscosities <50 MPa s. (Author)

A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests - 13342

Energy Technology Data Exchange (ETDEWEB)

Thien, Mike G. [Washington River Protection Solutions, LLC, P.O Box 850, Richland WA, 99352 (United States); Barnes, Steve M. [Waste Treatment Plant, 2435 Stevens Center Place, Richland WA 99354 (United States)

2013-07-01

The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described. (authors)

A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests - 13342

International Nuclear Information System (INIS)

Thien, Mike G.; Barnes, Steve M.

2013-01-01

The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described. (authors)

Waste disposal technology transfer matching requirement clusters for waste disposal facilities in China.

Science.gov (United States)

Dorn, Thomas; Nelles, Michael; Flamme, Sabine; Jinming, Cai

2012-11-01

Even though technology transfer has been part of development aid programmes for many decades, it has more often than not failed to come to fruition. One reason is the absence of simple guidelines or decision making tools that help operators or plant owners to decide on the most suitable technology to adopt. Practical suggestions for choosing the most suitable technology to combat a specific problem are hard to get and technology drawbacks are not sufficiently highlighted. Western counterparts in technology transfer or development projects often underestimate or don't sufficiently account for the high investment costs for the imported incineration plant; the differing nature of Chinese MSW; the need for trained manpower; and the need to treat flue gas, bunker leakage water, and ash, all of which contain highly toxic elements. This article sets out requirements for municipal solid waste disposal plant owner/operators in China as well as giving an attribute assessment for the prevalent waste disposal plant types in order to assist individual decision makers in their evaluation process for what plant type might be most suitable in a given situation. There is no 'best' plant for all needs and purposes, and requirement constellations rely on generalisations meaning they cannot be blindly applied, but an alignment of a type of plant to a type of owner or operator can realistically be achieved. To this end, a four-step approach is suggested and a technology matrix is set out to ease the choice of technology to transfer and avoid past errors. The four steps are (1) Identification of plant owner/operator requirement clusters; (2) Determination of different municipal solid waste (MSW) treatment plant attributes; (3) Development of a matrix matching requirement clusters to plant attributes; (4) Application of Quality Function Deployment Method to aid in technology localisation. The technology transfer matrices thus derived show significant performance differences between the

Railcar waste transfer system hydrostatic test report

International Nuclear Information System (INIS)

Ellingson, S.D.

1997-01-01

This Acceptance Test Report (ATR) documents for record purposes the field results, acceptance, and approvals of the completed acceptance test per HNF-SD-W417-ATP-001, ''Rail car Waste Transfer System Hydrostatic Test''. The test was completed and approved without any problems or exceptions

Physical sampling for site and waste characterization

International Nuclear Information System (INIS)

Bonnough, T.L.

1994-01-01

Physical sampling plays a basic role in site and waste characterization program effort. The term ''physical sampling'' used here means collecting tangible, physical samples of soil, water, air, waste streams, or other materials. The industry defines the term ''physical sampling'' broadly to include measurements of physical conditions such as temperature, wind conditions, and pH which are also often taken in a sample collection effort. Most environmental compliance actions are supported by the results of taking, recording, and analyzing physical samples and the measuring of physical conditions taken in association with sample collecting

W-026, transuranic waste restricted waste management (TRU RWM) glovebox operational test report

Energy Technology Data Exchange (ETDEWEB)

Leist, K.J.

1998-02-18

The TRU Waste/Restricted Waste Management (LLW/PWNP) Glovebox 401 is designed to accept and process waste from the Transuranic Process Glovebox 302. Waste is transferred to the glovebox via the Drath and Schraeder Bagless Transfer Port (DO-07401) on a transfer stand. The stand is removed with a hoist and the operator inspects the waste (with the aid of the Sampling and Treatment Director) to determine a course of action for each item. The waste is separated into compliant and non compliant. One Trip Port DO-07402A is designated as ``Compliant``and One Trip Port DO-07402B is designated as ``Non Compliant``. As the processing (inspection, bar coding, sampling and treatment) of the transferred items takes place, residue is placed in the appropriate One Trip port. The status of the waste items is tracked by the Data Management System (DMS) via the Plant Control System (PCS) barcode interface. As an item is moved for sampling or storage or it`s state altered by treatment, the Operator will track an items location using a portable barcode reader and entry any required data on the DMS console. The Operational Test Procedure (OTP) will perform evolutions (described here) using the Plant Operating Procedures (POP) in order to verify that they are sufficient and accurate for controlled glovebox operation.

W-026, transuranic waste restricted waste management (TRU RWM) glovebox operational test report

International Nuclear Information System (INIS)

Leist, K.J.

1998-01-01

The TRU Waste/Restricted Waste Management (LLW/PWNP) Glovebox 401 is designed to accept and process waste from the Transuranic Process Glovebox 302. Waste is transferred to the glovebox via the Drath and Schraeder Bagless Transfer Port (DO-07401) on a transfer stand. The stand is removed with a hoist and the operator inspects the waste (with the aid of the Sampling and Treatment Director) to determine a course of action for each item. The waste is separated into compliant and non compliant. One Trip Port DO-07402A is designated as ''Compliant''and One Trip Port DO-07402B is designated as ''Non Compliant''. As the processing (inspection, bar coding, sampling and treatment) of the transferred items takes place, residue is placed in the appropriate One Trip port. The status of the waste items is tracked by the Data Management System (DMS) via the Plant Control System (PCS) barcode interface. As an item is moved for sampling or storage or it's state altered by treatment, the Operator will track an items location using a portable barcode reader and entry any required data on the DMS console. The Operational Test Procedure (OTP) will perform evolutions (described here) using the Plant Operating Procedures (POP) in order to verify that they are sufficient and accurate for controlled glovebox operation

WRAP Module 1 sampling strategy and waste characterization alternatives study

Energy Technology Data Exchange (ETDEWEB)

Bergeson, C.L.

1994-09-30

The Waste Receiving and Processing Module 1 Facility is designed to examine, process, certify, and ship drums and boxes of solid wastes that have a surface dose equivalent of less than 200 mrem/h. These wastes will include low-level and transuranic wastes that are retrievably stored in the 200 Area burial grounds and facilities in addition to newly generated wastes. Certification of retrievably stored wastes processing in WRAP 1 is required to meet the waste acceptance criteria for onsite treatment and disposal of low-level waste and mixed low-level waste and the Waste Isolation Pilot Plant Waste Acceptance Criteria for the disposal of TRU waste. In addition, these wastes will need to be certified for packaging in TRUPACT-II shipping containers. Characterization of the retrievably stored waste is needed to support the certification process. Characterization data will be obtained from historical records, process knowledge, nondestructive examination nondestructive assay, visual inspection of the waste, head-gas sampling, and analysis of samples taken from the waste containers. Sample characterization refers to the method or methods that are used to test waste samples for specific analytes. The focus of this study is the sample characterization needed to accurately identify the hazardous and radioactive constituents present in the retrieved wastes that will be processed in WRAP 1. In addition, some sampling and characterization will be required to support NDA calculations and to provide an over-check for the characterization of newly generated wastes. This study results in the baseline definition of WRAP 1 sampling and analysis requirements and identifies alternative methods to meet these requirements in an efficient and economical manner.

WRAP Module 1 sampling strategy and waste characterization alternatives study

International Nuclear Information System (INIS)

Bergeson, C.L.

1994-01-01

The Waste Receiving and Processing Module 1 Facility is designed to examine, process, certify, and ship drums and boxes of solid wastes that have a surface dose equivalent of less than 200 mrem/h. These wastes will include low-level and transuranic wastes that are retrievably stored in the 200 Area burial grounds and facilities in addition to newly generated wastes. Certification of retrievably stored wastes processing in WRAP 1 is required to meet the waste acceptance criteria for onsite treatment and disposal of low-level waste and mixed low-level waste and the Waste Isolation Pilot Plant Waste Acceptance Criteria for the disposal of TRU waste. In addition, these wastes will need to be certified for packaging in TRUPACT-II shipping containers. Characterization of the retrievably stored waste is needed to support the certification process. Characterization data will be obtained from historical records, process knowledge, nondestructive examination nondestructive assay, visual inspection of the waste, head-gas sampling, and analysis of samples taken from the waste containers. Sample characterization refers to the method or methods that are used to test waste samples for specific analytes. The focus of this study is the sample characterization needed to accurately identify the hazardous and radioactive constituents present in the retrieved wastes that will be processed in WRAP 1. In addition, some sampling and characterization will be required to support NDA calculations and to provide an over-check for the characterization of newly generated wastes. This study results in the baseline definition of WRAP 1 sampling and analysis requirements and identifies alternative methods to meet these requirements in an efficient and economical manner

Waste disposal technology transfer matching requirement clusters for waste disposal facilities in China

International Nuclear Information System (INIS)

Dorn, Thomas; Nelles, Michael; Flamme, Sabine; Jinming, Cai

2012-01-01

Highlights: ► We outline the differences of Chinese MSW characteristics from Western MSW. ► We model the requirements of four clusters of plant owner/operators in China. ► We examine the best technology fit for these requirements via a matrix. ► Variance in waste input affects result more than training and costs. ► For China technology adaptation and localisation could become push, not pull factors. - Abstract: Even though technology transfer has been part of development aid programmes for many decades, it has more often than not failed to come to fruition. One reason is the absence of simple guidelines or decision making tools that help operators or plant owners to decide on the most suitable technology to adopt. Practical suggestions for choosing the most suitable technology to combat a specific problem are hard to get and technology drawbacks are not sufficiently highlighted. Western counterparts in technology transfer or development projects often underestimate or don’t sufficiently account for the high investment costs for the imported incineration plant; the differing nature of Chinese MSW; the need for trained manpower; and the need to treat flue gas, bunker leakage water, and ash, all of which contain highly toxic elements. This article sets out requirements for municipal solid waste disposal plant owner/operators in China as well as giving an attribute assessment for the prevalent waste disposal plant types in order to assist individual decision makers in their evaluation process for what plant type might be most suitable in a given situation. There is no ‘best’ plant for all needs and purposes, and requirement constellations rely on generalisations meaning they cannot be blindly applied, but an alignment of a type of plant to a type of owner or operator can realistically be achieved. To this end, a four-step approach is suggested and a technology matrix is set out to ease the choice of technology to transfer and avoid past errors. The four

Reconnaissance survey of the intermediate-level liquid waste transfer line between X-10 and the hydrofracture site

International Nuclear Information System (INIS)

Duguid, J.O.; Sealand, O.M.

1975-08-01

Two leakage points on an intermediate-level liquid waste line were located. The waste line is used periodically to transfer waste between X-10 and the hydrofracture site. The first leak occurred prior to this survey and had been repaired, but no contaminated soil had been removed. The second leak resulted in soil contamination that was more intense than at the first leak. Analyses of soil samples taken from both locations are given in this report. Groundwater data indicate the effectiveness of the removal of the contaminated material from leak two. 1 ref., 5 figs., 3 tabs

Reconnaissance survey of the intermediate level liquid waste transfer line between X-10 and the hydrofracture site

International Nuclear Information System (INIS)

Duguid, J.O.; Sealand, O.M.

1975-08-01

Two leakage points on an intermediate-level liquid waste line were located. The waste line is used periodically to transfer waste between X-10 and the hydrofracture site. The first leak had occurred prior to this survey and had been repaired. However, no contaminated soil had been removed. The second leak had not been discovered previously and soil contamination in this area was more intense than at the first leak. Analyses of soil samples taken from both locations are given in this report. Groundwater data that indicate the effectiveness of the removal of the contaminated material from leak two are presented. (U.S.)

Nitrate Waste Treatment Sampling and Analysis Plan

Energy Technology Data Exchange (ETDEWEB)

Vigil-Holterman, Luciana R. [Los Alamos National Laboratory; Martinez, Patrick Thomas [Los Alamos National Laboratory; Garcia, Terrence Kerwin [Los Alamos National Laboratory

2017-07-05

This plan is designed to outline the collection and analysis of nitrate salt-bearing waste samples required by the New Mexico Environment Department- Hazardous Waste Bureau in the Los Alamos National Laboratory (LANL) Hazardous Waste Facility Permit (Permit).

Research on Layout Optimization of Urban Circle Solid Waste Transfer and Disposal Stations

OpenAIRE

Xuhui Li; Gangyan Li; Guowen Sun; Huiping Shi; Bao’an Yang

2013-01-01

Based on the Systematic Layout Planning theory and the analysis of transfer stations’ technological processes, a layout optimization model for solid waste transfer and disposal stations was made. The operating units’ layout of the solid waste transfer and disposal stations was simulated and optimized using the genetic algorithm, which could achieve reasonable technological processes, the smallest floor space and the lowest construction cost. The simulation result can also direct t...

Comparison of Waste Feed Delivery Small Scale Mixing Demonstration Simulant to Hanford Waste

Energy Technology Data Exchange (ETDEWEB)

Wells, Beric E.; Gauglitz, Phillip A.; Rector, David R.

2012-07-10

The Hanford double-shell tank (DST) system provides the staging location for waste that will be transferred to the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Specific WTP acceptance criteria for waste feed delivery describe the physical and chemical characteristics of the waste that must be met before the waste is transferred from the DSTs to the WTP. One of the more challenging requirements relates to the sampling and characterization of the undissolved solids (UDS) in a waste feed DST because the waste contains solid particles that settle and their concentration and relative proportion can change during the transfer of the waste in individual batches. A key uncertainty in the waste feed delivery system is the potential variation in UDS transferred in individual batches in comparison to an initial sample used for evaluating the acceptance criteria. To address this uncertainty, a number of small-scale mixing tests have been conducted as part of Washington River Protection Solutions' Small Scale Mixing Demonstration (SSMD) project to determine the performance of the DST mixing and sampling systems. A series of these tests have used a five-part simulant composed of particles of different size and density and designed to be equal or more challenging than AY-102 waste. This five-part simulant, however, has not been compared with the broad range of Hanford waste, and thus there is an additional uncertainty that this simulant may not be as challenging as the most difficult Hanford waste. The purpose of this study is to quantify how the current five-part simulant compares to all of the Hanford sludge waste, and to suggest alternate simulants that could be tested to reduce the uncertainty in applying the current testing results to potentially more challenging wastes.

Double-shell tank waste transfer facilities integrity assessment plan

International Nuclear Information System (INIS)

Hundal, T.S.

1998-01-01

This document presents the integrity assessment plan for the existing double-shell tank waste transfer facilities system in the 200 East and 200 West Areas of Hanford Site. This plan identifies and proposes the integrity assessment elements and techniques to be performed for each facility. The integrity assessments of existing tank systems that stores or treats dangerous waste is required to be performed to be in compliance with the Washington State Department of Ecology Dangerous Waste Regulations, Washington Administrative Code WAC-173-303-640 requirements

Physical sampling for site and waste characterization

International Nuclear Information System (INIS)

Bonnough, T.L.

1996-01-01

Physical sampling plays a basic role in high-level radioactive waste management program effort. The term ''physical sampling'' used here means collecting tangible, physical samples of soil, water, air, waste streams, or other materials. The industry defines the term ''physical sampling'' broadly to include measurements of physical conditions such as temperature, wind conditions, and pH, which are also often taken in a sample collection effort. Most environmental compliance actions are supported by the results of taking, recording, and analyzing physical samples and the measurements of physical conditions taken in association with sample collecting. Therefore, the when and how to take samples is needed to be known and planned

40 CFR 761.269 - Sampling liquid PCB remediation waste.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 30 2010-07-01 2010-07-01 false Sampling liquid PCB remediation waste..., AND USE PROHIBITIONS Cleanup Site Characterization Sampling for PCB Remediation Waste in Accordance with § 761.61(a)(2) § 761.269 Sampling liquid PCB remediation waste. (a) If the liquid is single phase...

Site-specific waste management instruction for the field sampling organization

International Nuclear Information System (INIS)

Bryant, D.L.

1997-01-01

The Site-Specific Waste Management Instruction (SSWMI) provides guidance for the management of waste generated from field-sampling activities performed by the Environment Restoration Contractor (ERC) Sampling Organization that are not managed as part of a project SSWMI. Generally, the waste is unused preserved groundwater trip blanks, used and expired calibration solutions, and other similar waste that cannot be returned to an ERC project for disposal. The specific waste streams addressed by this SSWMI are identified in Section 2.0. This SSWMI was prepared in accordance with BHI-EE-02, Environmental Requirements. Waste generated from field sample collection activities should be returned to the project and managed in accordance with the applicable project-specific SSWMI whenever possible. However, returning all field sample collection and associated waste to a project for disposal may not always be practical or cost effective. Therefore, the ERC field sampling organization must manage and arrange to dispose of the waste using the (Bechtel Hanford, Inc. [BHI]) Field Support Waste Management (FSWM) services. This SSWMI addresses those waste streams that are the responsibility of the field sampling organization to manage and make arrangements for disposal

DOE methods for evaluating environmental and waste management samples

International Nuclear Information System (INIS)

Goheen, S.C.; McCulloch, M.; Thomas, B.L.; Riley, R.G.; Sklarew, D.S.; Mong, G.M.; Fadeff, S.K.

1993-03-01

DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods) provides applicable methods in use by. the US Department of Energy (DOE) laboratories for sampling and analyzing constituents of waste and environmental samples. The development of DOE Methods is supported by the Laboratory Management Division (LMD) of the DOE. This document contains chapters and methods that are proposed for use in evaluating components of DOE environmental and waste management samples. DOE Methods is a resource intended to support sampling and analytical activities that will aid in defining the type and breadth of contamination and thus determine the extent of environmental restoration or waste management actions needed, as defined by the DOE, the US Environmental Protection Agency (EPA), or others

DOE methods for evaluating environmental and waste management samples.

Energy Technology Data Exchange (ETDEWEB)

Goheen, S C; McCulloch, M; Thomas, B L; Riley, R G; Sklarew, D S; Mong, G M; Fadeff, S K [eds.; Pacific Northwest Lab., Richland, WA (United States)

1994-04-01

DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods) provides applicable methods in use by. the US Department of Energy (DOE) laboratories for sampling and analyzing constituents of waste and environmental samples. The development of DOE Methods is supported by the Laboratory Management Division (LMD) of the DOE. This document contains chapters and methods that are proposed for use in evaluating components of DOE environmental and waste management samples. DOE Methods is a resource intended to support sampling and analytical activities that will aid in defining the type and breadth of contamination and thus determine the extent of environmental restoration or waste management actions needed, as defined by the DOE, the US Environmental Protection Agency (EPA), or others.

Los Alamos National Laboratory TRU waste sampling projects

International Nuclear Information System (INIS)

Yeamans, D.; Rogers, P.; Mroz, E.

1997-01-01

The Los Alamos National Laboratory (LANL) has begun characterizing transuranic (TRU) waste in order to comply with New Mexico regulations, and to prepare the waste for shipment and disposal at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico. Sampling consists of removing some head space gas from each drum, removing a core from a few drums of each homogeneous waste stream, and visually characterizing a few drums from each heterogeneous waste stream. The gases are analyzed by GC/MS, and the cores are analyzed for VOC's and SVOC's by GC/MS and for metals by AA or AE spectroscopy. The sampling and examination projects are conducted in accordance with the ''DOE TRU Waste Quality Assurance Program Plan'' (QAPP) and the ''LANL TRU Waste Quality Assurance Project Plan,'' (QAPjP), guaranteeing that the data meet the needs of both the Carlsbad Area Office (CAO) of DOE and the ''WIPP Waste Acceptance Criteria, Rev. 5,'' (WAC)

Tem holder for sample transfer under reaction conditions

DEFF Research Database (Denmark)

Damsgaard, Christian Danvad; Zandbergen, Henny W.; Wagner, Jakob Birkedal

Environmental transmission electron microscopy (ETEM) studies are usually performed using conventional sample holders in a dedicated ETEM [1] or in a traditional TEM by use of a dedicated high-pressure cell sample holder [2]. In both cases, the setup defines the conditions regarding gas, pressure......]. Furthermore, dedicated transfer holders have been used to transfer catalyst samples between reactor set-ups and TEM at room temperature in inert atmosphere [5]. To take the full advantage of complementary in situ techniques, transfer under reactions conditions is essential. This study introduces the in situ...... transfer concept by use of a dedicated TEM transfer holder capable of enclosing the sample in a gaseous environment at temperatures up to approx. 900C. By oxidation and reduction experiments of Cu nanoparticles it is shown possible to keep the reaction conditions during transfer outside the microscope...

Heat transfer analyses for grout disposal of radioactive double-shell slurry and customer wastes

International Nuclear Information System (INIS)

Robinson, S.M.; Gilliam, T.M.; McDaniel, E.W.

1987-04-01

Grout immobilization is being considered by Rockwell Hanford Operations (Rockwell Hanford) as a permanent disposal method for several radioactive waste streams. These include disposal of customer and double-shell slurry wastes in earthen trenches and in single-shell underground waste storage tanks. Heat transfer studies have previously been made to determine the maximum heat loading for grout disposal of various wastes under similar conditions, but a sensitivity analysis of temperature profiles to input parameters was needed. This document presents the results of heat transfer calculations for trenches containing grouted customer and double-shell slurry wastes and for in situ disposal of double-shell wastes in single-shell, domed concrete storage tanks. It discusses the conditions that lead to maximum grout temperatures of 250 0 F during the curing stage and 350 0 F thereafter and shows the dependence of these temperatures on input parameters such as soil and grout thermal conductivity, grout specific heat, waste loading, and disposal geometries. Transient heat transfer calculations were made using the HEATING6 computer code to predict temperature profiles in solidified low-level radioactive waste disposal scenarios at the Rockwell Hanford site. The calculations provide guidance for the development of safe, environmentally acceptable grout formulas for the Transportable Grout Facility. 11 refs

Transuranic waste characterization sampling and analysis plan

International Nuclear Information System (INIS)

1994-01-01

Los Alamos National Laboratory (the Laboratory) is located approximately 25 miles northwest of Santa Fe, New Mexico, situated on the Pajarito Plateau. Technical Area 54 (TA-54), one of the Laboratory's many technical areas, is a radioactive and hazardous waste management and disposal area located within the Laboratory's boundaries. The purpose of this transuranic waste characterization, sampling, and analysis plan (CSAP) is to provide a methodology for identifying, characterizing, and sampling approximately 25,000 containers of transuranic waste stored at Pads 1, 2, and 4, Dome 48, and the Fiberglass Reinforced Plywood Box Dome at TA-54, Area G, of the Laboratory. Transuranic waste currently stored at Area G was generated primarily from research and development activities, processing and recovery operations, and decontamination and decommissioning projects. This document was created to facilitate compliance with several regulatory requirements and program drivers that are relevant to waste management at the Laboratory, including concerns of the New Mexico Environment Department

76 FR 16538 - Solid Waste Rail Transfer Facilities

Science.gov (United States)

2011-03-24

... leaving in place the rules issued in 2009, which were drafted without any input from industry and other... discarded by residential dwellings, hotels, motels, and other similar permanent or temporary housing... notice concerning the acquisition shall include a statement that a solid waste rail transfer facility...

A destructive sample preparation method for radioactive waste characterization

International Nuclear Information System (INIS)

Olteanu, M.; Bucur, C.

2015-01-01

Acid digestion, using the microwave power, was applied for ''dissolution'' of different materials corresponding to the radioactive waste matrices resulted from a nuclear power plant operation, including exchange resin (cationic and mixed), concrete, paper, textile and activated charcoals. A small aliquot of solid sample (0.1-0.5g) was mixed with a known volume of digestion reagents (HNO3 67% - H2O2 30% or HNO3 67% - HCl 37%, with HF addition if the SiO2 was present in matrices) in a 100 ml PTFE vessel and it was mineralized using a Berghof digestion system, Speedwave 4. Starting from the manufacturer procedures, the technical parameters (temperature and mineralization time), the types and quantities of digestion reagents were optimized. After the mineralization process, the samples were transferred in centrifuge tubes, separated at 3500 rot/min and visually analysed. The obtained solutions were clear, without suspended or deposed materials and separated phases, ready for future separation processes of the ''difficult to measure'' radioisotopes. (authors)

A STRUCTURAL INTEGRITY EVALUATION OF THE TANK FARM WASTE TRANSFER SYSTEM

Energy Technology Data Exchange (ETDEWEB)

Wiersma, B.

2006-03-09

Radioactive supernate, salt, and/or sludge wastes (i.e., high level wastes) are confined in 49 underground storage tanks at the Savannah River Site (SRS). The waste is transported between tanks within and between the F and H area tank farms and other facilities on site via underground and a limited number of aboveground transfer lines. The Department of Energy - Savannah River Operations Office (DOE-SR) performed a comprehensive assessment of the structural integrity program for the Tank Farm waste transfer system at the SRS. This document addresses the following issues raised during the DOE assessment: (1) Inspections of failed or replaced transfer lines indicated that the wall thickness of some core and jacket piping is less than nominal; (2) No corrosion allowance is utilized in the transfer line structural qualification calculations. No basis for neglecting corrosion was provided in the calculations; (3) Wall loss due to erosion is not addressed in the transfer line structural qualification calculations; and (4) No basis is provided for neglecting intergranular stress corrosion cracking in the transfer line structural qualification calculations. The common theme in most of these issues is the need to assess the potential for occurrence of material degradation of the transfer line piping. The approach used to resolve these issues involved: (1) Review the design and specifications utilized to construct and fabricate the piping system; (2) Review degradation mechanisms for stainless steel and carbon steel and determine their relevance to the transfer line piping; (3) Review the transfer piping inspection data; (4) Life estimation calculations for the transfer lines; and (5) A Fitness-For-Service evaluation for one of the transfer line jackets. The evaluation concluded that the transfer line system piping has performed well for over fifty years. Although there have been instances of failures of the stainless steel core pipe during off-normal service, no significant

Determination of Erosion/Corrosion Rates in Hanford Tank Farms Radioactive Waste Transfer System Pipelines

International Nuclear Information System (INIS)

Washenfelder, D. J.; Girardot, C. L.; Wilson, E. R.; Page, J. A.; Engeman, J. K.; Gunter, J. R.; Johnson, J. M.; Baide, D. G.; Cooke, G. A.; Larson, J. D.; Castleberry, J. L.; Boomer, K. D.

2015-01-01

The twenty-eight double-shell underground radioactive waste storage tanks at the U. S. Department of Energy's Hanford Site near Richland, WA are interconnected by the Waste Transfer System network of buried steel encased pipelines and pipe jumpers in below-grade pits. The pipeline material is stainless steel or carbon steel in 51 mm to 152 mm (2 in. to 6 in.) sizes. The pipelines carry slurries ranging up to 20 volume percent solids and supernatants with less than one volume percent solids at velocities necessary to prevent settling. The pipelines, installed between 1976 and 2011, were originally intended to last until the 2028 completion of the double-shell tank storage mission. The mission has been subsequently extended. In 2010 the Tank Operating Contractor began a systematic evaluation of the Waste Transfer System pipeline conditions applying guidelines from API 579-1/ASME FFS-1 (2007), Fitness-For-Service. Between 2010 and 2014 Fitness-for-Service examinations of the Waste Transfer System pipeline materials, sizes, and components were completed. In parallel, waste throughput histories were prepared allowing side-by-side pipeline wall thinning rate comparisons between carbon and stainless steel, slurries and supernatants and throughput volumes. The work showed that for transfer volumes up to 6.1E+05 m"3 (161 million gallons), the highest throughput of any pipeline segment examined, there has been no detectable wall thinning in either stainless or carbon steel pipeline material regardless of waste fluid characteristics or throughput. The paper describes the field and laboratory evaluation methods used for the Fitness-for-Service examinations, the results of the examinations, and the data reduction methodologies used to support Hanford Waste Transfer System pipeline wall thinning conclusions.

Determination of Erosion/Corrosion Rates in Hanford Tank Farms Radioactive Waste Transfer System Pipelines

Energy Technology Data Exchange (ETDEWEB)

Washenfelder, D. J.; Girardot, C. L.; Wilson, E. R.; Page, J. A.; Engeman, J. K.; Gunter, J. R.; Johnson, J. M.; Baide, D. G.; Cooke, G. A.; Larson, J. D.; Castleberry, J. L.; Boomer, K. D.

2015-11-05

The twenty-eight double-shell underground radioactive waste storage tanks at the U. S. Department of Energy’s Hanford Site near Richland, WA are interconnected by the Waste Transfer System network of buried steel encased pipelines and pipe jumpers in below-grade pits. The pipeline material is stainless steel or carbon steel in 51 mm to 152 mm (2 in. to 6 in.) sizes. The pipelines carry slurries ranging up to 20 volume percent solids and supernatants with less than one volume percent solids at velocities necessary to prevent settling. The pipelines, installed between 1976 and 2011, were originally intended to last until the 2028 completion of the double-shell tank storage mission. The mission has been subsequently extended. In 2010 the Tank Operating Contractor began a systematic evaluation of the Waste Transfer System pipeline conditions applying guidelines from API 579-1/ASME FFS-1 (2007), Fitness-For-Service. Between 2010 and 2014 Fitness-for-Service examinations of the Waste Transfer System pipeline materials, sizes, and components were completed. In parallel, waste throughput histories were prepared allowing side-by-side pipeline wall thinning rate comparisons between carbon and stainless steel, slurries and supernatants and throughput volumes. The work showed that for transfer volumes up to 6.1E+05 m³ (161 million gallons), the highest throughput of any pipeline segment examined, there has been no detectable wall thinning in either stainless or carbon steel pipeline material regardless of waste fluid characteristics or throughput. The paper describes the field and laboratory evaluation methods used for the Fitness-for-Service examinations, the results of the examinations, and the data reduction methodologies used to support Hanford Waste Transfer System pipeline wall thinning conclusions.

Waste sampling and characterization facility (WSCF)

International Nuclear Information System (INIS)

1994-10-01

The Waste Sampling and Characterization Facility (WSCF) complex consists of the main structure (WSCF) and four support structures located in the 600 Area of the Hanford site east of the 200 West area and south of the Hanford Meterology Station. WSCF is to be used for low level sample analysis, less than 2 mRem. The Laboratory features state-of-the-art analytical and low level radiological counting equipment for gaseous, soil, and liquid sample analysis. In particular, this facility is to be used to perform Resource Conservation and Recovery Act (RCRA) of 1976 and Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) of 1980 sample analysis in accordance with U.S. Environmental Protection Agency Protocols, room air and stack monitoring sample analysis, waste water treatment process support, and contractor laboratory quality assurance checks. The samples to be analyzed contain very low concentrations of radioisotopes. The main reason that WSCF is considered a Nuclear Facility is due to the storage of samples at the facility. This maintenance Implementation Plan has been developed for maintenace functions associate with the WSCF

Operating test report for project W-417, T-plant steam removal upgrade, waste transfer portion

International Nuclear Information System (INIS)

Myers, N.K.

1997-01-01

This Operating Test Report (OTR) documents the performance results of the Operating Test Procedure HNF-SD-W417-OTP-001 that provides steps to test the waste transfer system installed in the 221-T Canyon under project W-417. Recent modifications have been performed on the T Plant Rail Car Waste Transfer System. This Operating Test Procedure (OTP) will document the satisfactory operation of the 221-T Rail Car Waste Transfer System modified by project W-417. Project W-417 installed a pump in Tank 5-7 to replace the steam jets used for transferring liquid waste. This testing is required to verify that operational requirements of the modified transfer system have been met. Figure 2 and 3 shows the new and existing system to be tested. The scope of this testing includes the submersible air driven pump operation in Tank 5-7, liquid waste transfer operation from Tank 5-7 to rail car (HO-IOH-3663 or HO-IOH-3664), associated line flushing, and the operation of the flow meter. This testing is designed to demonstrate the satisfactory operation-of the transfer line at normal operating conditions and proper functioning of instruments. Favorable results will support continued use of this system for liquid waste transfer. The Functional Design Criteria for this system requires a transfer flow rate of 40 gallons per minute (GPM). To establish these conditions the pump will be supplied up to 90 psi air pressure from the existing air system routed in the canyon. An air regulator valve will regulate the air pressure. Tank capacity and operating ranges are the following: Tank No. Capacity (gal) Operating Range (gal) 5-7 10,046 0 8040 (80%) Rail car (HO-IOH-3663 HO-IOH-3664) 097219,157 Existing Tank level instrumentation, rail car level detection, and pressure indicators will be utilized for acceptance/rejection Criteria. The flow meter will be verified for accuracy against the Tank 5-7 level indicator. The level indicator is accurate to within 2.2 %. This will be for information only

Results of sampling the contents of the liquid low-level waste evaporator feed tank W-22 at ORNL

International Nuclear Information System (INIS)

Sears, M.B.

1996-09-01

This report summarizes the results of the fall 1994 sampling of the contents of the liquid low- level waste (LLLW) tank W-22 at the Oak Ridge National Laboratory (ORNL). Tank W-22 is the central collection and holding tank for LLLW at ORNL before the waste is transferred to the evaporators. Samples of the tank liquid and sludge were analyzed to determine (1) the major chemical constituents, (2) the principal radionuclides, (3) the metals listed on the U.S. Environmental Protection Agency (EPA) Contract Laboratory Program Inorganic Target Analyte List, (4) organic compounds, and (5) some physical properties. The organic chemical characterization consisted of the determinations of the EPA Contract Laboratory Program Target Compound List semivolatile compounds, pesticides, and polychlorinated biphenyls (PCBs). Water-soluble volatile organic compounds were also determined. Information provided in this report forms part of the technical basis in support of (1) waste management for the active LLLW system and (2) planning for the treatment and disposal of the waste

Statistical sampling applied to the radiological characterization of historical waste

Directory of Open Access Journals (Sweden)

Zaffora Biagio

2016-01-01

Full Text Available The evaluation of the activity of radionuclides in radioactive waste is required for its disposal in final repositories. Easy-to-measure nuclides, like γ-emitters and high-energy X-rays, can be measured via non-destructive nuclear techniques from outside a waste package. Some radionuclides are difficult-to-measure (DTM from outside a package because they are α- or β-emitters. The present article discusses the application of linear regression, scaling factors (SF and the so-called “mean activity method” to estimate the activity of DTM nuclides on metallic waste produced at the European Organization for Nuclear Research (CERN. Various statistical sampling techniques including simple random sampling, systematic sampling, stratified and authoritative sampling are described and applied to 2 waste populations of activated copper cables. The bootstrap is introduced as a tool to estimate average activities and standard errors in waste characterization. The analysis of the DTM Ni-63 is used as an example. Experimental and theoretical values of SFs are calculated and compared. Guidelines for sampling historical waste using probabilistic and non-probabilistic sampling are finally given.

Results of Waste Transfer and Back-Dilution in Tanks 241-SY-101 and 241-SY-102

International Nuclear Information System (INIS)

Mahoney, L.A.; Antoniak, Z.I.; Barton, W.B.; Conner, J.M.; Kirch, N.W.; Stewart, C.W.; Wells, B.E.

2000-01-01

This report chronicles the process of remediation of the flammable gas hazard in Tank 241-SY-101 (SY-101) by waste transfer and back-dilution from December 18, 1999 through April 2, 2000. A brief history is given of the development of the flammable gas retention and release hazard in this tank, and the transfer and dilution systems are outlined. A detailed narrative of each of the three transfer and dilution campaigns is given to provide structure for the balance of the report. Details of the behavior of specific data are then described, including the effect of transfer and dilution on the waste levels in Tanks SY-101 and SY-102, data from strain gauges on equipment suspended from the tank dome, changes in waste configuration as inferred from neutron and gamma logs, headspace gas concentrations, waste temperatures, and the mixerpump operating performance. Operating data and performance of the transfer pump in SY-101 are also discussed

Effects of stabilizers on the heat transfer characteristics of a nuclear waste canister

International Nuclear Information System (INIS)

Vafai, K.; Ettefagh, J.

1986-07-01

This report summarizes the feasibility and the effectiveness of using stabilizers (internal metal structural components) to augment the heat transfer characteristics of a nuclear waste canister. The problem was modeled as a transient two-dimensional heat transfer in two physical domains - the stabilizer and the wedge (a 30-degree-angle canister segment), which includes the heat-producing spent-fuel rods. This problem is solved by a simultaneous and interrelated numerical investigation of the two domains in cartesian and polar coordinate systems. The numerical investigations were performed for three cases. In the first case, conduction was assumed to be the dominant mechanism for heat transfer. The second case assumed that radiation was the dominant mechanism, and in the third case both radiation and conduction were considered as mechanisms of heat transfer. The results show that for typical conditions in a waste package design, the stabilizers are quite effective in reducing the overall temperature in a waste canister. Furthermore, the results show that increasing the stabilizer thickness over the thickness specified in the present design has a negligible effect on the temperature distribution in the canister. Finally, the presence of the stabilizers was found to shift the location of the peak temperature areas in the waste canister

Sampling and analysis strategies to support waste form qualification

International Nuclear Information System (INIS)

Westsik, J.H. Jr.; Pulsipher, B.A.; Eggett, D.L.; Kuhn, W.L.

1989-04-01

As part of the waste acceptance process, waste form producers will be required to (1) demonstrate that their glass waste form will meet minimum specifications, (2) show that the process can be controlled to consistently produce an acceptable waste form, and (3) provide documentation that the waste form produced meets specifications. Key to the success of these endeavors is adequate sampling and chemical and radiochemical analyses of the waste streams from the waste tanks through the process to the final glass product. This paper suggests sampling and analysis strategies for meeting specific statistical objectives of (1) detection of compositions outside specification limits, (2) prediction of final glass product composition, and (3) estimation of composition in process vessels for both reporting and guiding succeeding process steps. 2 refs., 1 fig., 3 tabs

EVALUATION OF FROST HEAVE ON WASTE TRANSFER LINES WITH SHALLOW DEPTHS IN DST (DOUBLE SHELL TANK) FARMS

Energy Technology Data Exchange (ETDEWEB)

HAQ MA

2009-05-12

The purpose of this document is to evaluate the effect of frost heave on waste transfer lines with shallow depths in DST farms. Because of the insulation, well compacted sandy material around waste transfer lines, the type of sandy and gravel soil, and relatively low precipitation at Hanford site, it is concluded that waste transfer lines with one foot of soil covers (sandy cushion material and insulation) are not expected to undergo frost heave damaging effects.

Production of a High-Level Waste Glass from Hanford Waste Samples

International Nuclear Information System (INIS)

Crawford, C.L.; Farrara, D.M.; Ha, B.C.; Bibler, N.E.

1998-09-01

The HLW glass was produced from a HLW sludge slurry (Envelope D Waste), eluate waste streams containing high levels of Cs-137 and Tc-99, solids containing both Sr-90 and transuranics (TRU), and glass-forming chemicals. The eluates and Sr-90/TRU solids were obtained from ion-exchange and precipitation pretreatments, respectively, of other Hanford supernate samples (Envelopes A, B and C Waste). The glass was vitrified by mixing the different waste streams with glass-forming chemicals in platinum/gold crucibles and heating the mixture to 1150 degree C. Resulting glass analyses indicated that the HLW glass waste form composition was close to the target composition. The targeted waste loading of Envelope D sludge solids in the HLW glass was 30.7 wt percent, exclusive of Na and Si oxides. Condensate samples from the off-gas condenser and off-gas dry-ice trap indicated that very little of the radionuclides were volatilized during vitrification. Microstructure analysis of the HLW glass using Scanning Electron Microscopy (SEM) and Energy Dispersive X-Ray Analysis (EDAX) showed what appeared to be iron spinel in the HLW glass. Further X-Ray Diffraction (XRD) analysis confirmed the presence of nickel spinel trevorite (NiFe2O4). These crystals did not degrade the leaching characteristics of the glass. The HLW glass waste form passed leach tests that included a standard 90 degree C Product Consistency Test (PCT) and a modified version of the United States Environmental Protection Agency Toxicity Characteristic Leaching Procedure (TCLP)

Critique of Hanford Waste Vitrification Plant off-gas sampling requirements

International Nuclear Information System (INIS)

Goles, R.W.

1996-03-01

Off-gas sampling and monitoring activities needed to support operations safety, process control, waste form qualification, and environmental protection requirements of the Hanford Waste Vitrification Plant (HWVP) have been evaluated. The locations of necessary sampling sites have been identified on the basis of plant requirements, and the applicability of Defense Waste Processing Facility (DWPF) reference sampling equipment to these HWVP requirements has been assessed for all sampling sites. Equipment deficiencies, if present, have been described and the bases for modifications and/or alternative approaches have been developed

Impact of cementitious materials decalcification on transfer properties: application to radioactive waste deep repository

International Nuclear Information System (INIS)

Perlot, C.

2005-09-01

Cementitious materials have been selected to compose the engineering barrier system (EBS) of the French radioactive waste deep repository, because of concrete physico-chemical properties: the hydrates of the cementitious matrix and the pH of the pore solution contribute to radionuclides retention; furthermore the compactness of these materials limits elements transport. The confinement capacity of the system has to be assessed while a period at least equivalent to waste activity (up to 100.000 years). His durability was sustained by the evolution of transfer properties in accordance with cementitious materials decalcification, alteration that expresses structure long-term behavior. Then, two degradation modes were carried out, taking into account the different physical and chemical solicitations imposed by the host formation. The first mode, a static one, was an accelerated decalcification test using nitrate ammonium solution. It replicates the EBS alteration dues to underground water. Degradation kinetic was estimated by the amount of calcium leached and the measurement of the calcium hydroxide dissolution front. To evaluate the decalcification impact, samples were characterized before and after degradation in term of microstructure (porosity, pores size distribution) and of transfer properties (diffusivity, gas and water permeability). The influence of cement nature (ordinary Portland cement, blended cement) and aggregates type (lime or siliceous) was observed: experiments were repeated on different mortars mixes. On this occasion, an essential reflection on this test metrology was led. The second mode, a dynamical degradation, was performed with an environmental permeameter. It recreates the EBS solicitations ensured during the re-saturation period, distinguished by the hydraulic pressure imposed by the geologic layer and the waste exothermicity. This apparatus, based on triaxial cell functioning, allows applying on samples pressure drop between 2 and 10 MPa and

Design compliance matrix waste sample container filling system for nested, fixed-depth sampling system

International Nuclear Information System (INIS)

BOGER, R.M.

1999-01-01

This design compliance matrix document provides specific design related functional characteristics, constraints, and requirements for the container filling system that is part of the nested, fixed-depth sampling system. This document addresses performance, external interfaces, ALARA, Authorization Basis, environmental and design code requirements for the container filling system. The container filling system will interface with the waste stream from the fluidic pumping channels of the nested, fixed-depth sampling system and will fill containers with waste that meet the Resource Conservation and Recovery Act (RCRA) criteria for waste that contains volatile and semi-volatile organic materials. The specifications for the nested, fixed-depth sampling system are described in a Level 2 Specification document (HNF-3483, Rev. 1). The basis for this design compliance matrix document is the Tank Waste Remediation System (TWRS) desk instructions for design Compliance matrix documents (PI-CP-008-00, Rev. 0)

Mass Transfer Model for a Breached Waste Package

International Nuclear Information System (INIS)

Hsu, C.; McClure, J.

2004-01-01

The degradation of waste packages, which are used for the disposal of spent nuclear fuel in the repository, can result in configurations that may increase the probability of criticality. A mass transfer model is developed for a breached waste package to account for the entrainment of insoluble particles. In combination with radionuclide decay, soluble advection, and colloidal transport, a complete mass balance of nuclides in the waste package becomes available. The entrainment equations are derived from dimensionless parameters such as drag coefficient and Reynolds number and based on the assumption that insoluble particles are subjected to buoyant force, gravitational force, and drag force only. Particle size distributions are utilized to calculate entrainment concentration along with geochemistry model abstraction to calculate soluble concentration, and colloid model abstraction to calculate colloid concentration and radionuclide sorption. Results are compared with base case geochemistry model, which only considers soluble advection loss

Evaluation of the Validity of three Criteria for Sampling and Analyzing DST Wastes in Support of Waste Feed Delivery

International Nuclear Information System (INIS)

BOGER, R.M.

2000-01-01

This document summarizes the analysis of 3 basic criteria for the sampling systems that will provide waste validation samples of tank waste feeds prior to delivery to the waste treatment and immobilization plant where the wastes will be converted to glass forms. The assessed criteria includes sampling through a 4-inch riser, sampling while a mixer pump is operating, and the deployment of an at-tank analysis system. The assessment, based on the Phase I, 3S6 waste feed scenario, indicated that for high level waste, sampling through a 4-inch riser is not required but sampling while mixer pumps are operating will be required. For low activity waste, sampling through a 4-inch riser will be required but sampling while mixer pumps are operating is not required. The assessment indicated that an at-tank analysis system to provide tank mixing/settling (homogeneity) status is not needed since the number of tanks providing LAW feed was expanded and the payment basis in the original privatization contract has been modified

Estimating waste disposal quantities from raw waste samples

International Nuclear Information System (INIS)

Negin, C.A.; Urland, C.S.; Hitz, C.G.; GPU Nuclear Corp., Middletown, PA)

1985-01-01

Estimating the disposal quantity of waste resulting from stabilization of radioactive sludge is complex because of the many factors relating to sample analysis results, radioactive decay, allowable disposal concentrations, and options for disposal containers. To facilitate this estimation, a microcomputer spread sheet template was created. The spread sheet has saved considerable engineering hours. 1 fig., 3 tabs

Implementation of the waste management transfer act. Requirements from a regulatory point of view

International Nuclear Information System (INIS)

Mueller-Dehn, Christian

2017-01-01

In future in Germany, the state will be responsible for financing and handling the interim and final storage of radioactive waste from nuclear power plants. With regard to interim storage, this objective is achieved with the provisions of the Waste Management Transfer Act. Regulatory implementation is based on these regulations. BGZ Gesellschaft fuer Zwischenlager mbH is responsible for interim storage on behalf of the Federal Government. Simultaneously with the transfer of interim storage facilities to BGZ a legal transfer of approval is carried out. Insofar as there is a technical, organisational or personnel conjunction with the nuclear power plant operation, which continues to exist beyond this deadline and is relevant for regulatory purposes, a regulation is made via a service contract with the BGZ. This ensures compliance with the licensing regulations. Irradiated fuel assemblies and the waste from reprocessing can be handed over to BGZ from 1 January 2019 onwards and waste with negligible heat generation can be disposed of as of the determination of their proper packaging.

Analysis of Waste Isolation Pilot Plant Samples: Integrated Summary Report

Energy Technology Data Exchange (ETDEWEB)

Britt, Phillip F [ORNL

2015-03-01

Analysis of Waste Isolation Pilot Plant Samples: Integrated Summary Report. Summaries of conclusions, analytical processes, and analytical results. Analysis of samples taken from the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico in support of the WIPP Technical Assessment Team (TAT) activities to determine to the extent feasible the mechanisms and chemical reactions that may have resulted in the breach of at least one waste drum and release of waste material in WIPP Panel 7 Room 7 on February 14, 2014. This report integrates and summarizes the results contained in three separate reports, described below, and draws conclusions based on those results. Chemical and Radiochemical Analyses of WIPP Samples R-15 C5 SWB and R16 C-4 Lip; PNNL-24003, Pacific Northwest National Laboratory, December 2014 Analysis of Waste Isolation Pilot Plant (WIPP) Underground and MgO Samples by the Savannah River National Laboratory (SRNL); SRNL-STI-2014-00617; Savannah River National Laboratory, December 2014 Report for WIPP UG Sample #3, R15C5 (9/3/14); LLNL-TR-667015; Lawrence Livermore National Laboratory, January 2015 This report is also contained in the Waste Isolation Pilot Plant Technical Assessment Team Report; SRNL-RP-2015-01198; Savannah River National Laboratory, March 17, 2015, as Appendix C: Analysis Integrated Summary Report.

Waste package transfer, emplacement and retrievability in the French deep geological repository

Energy Technology Data Exchange (ETDEWEB)

Roulet, Alain; Delort, Daniel; Herve, Jean Francois; Bosgiraud, Jean Michel; Guenin, Jean Jacques [Technical Department ANDRA (France)

2009-06-15

Safe, reliable and reversible handling of waste is a significant issue related to the design and safety assessment of deep geological repository in France. The first step taken was to study various waste handling solutions. ANDRA also decided to fabricate and demonstrate industrial scale handling equipment for HLW (since 2003) and for ILW-LL wastes (since 2008). We will review the main equipment developed for the transfer process in the repository, for both types of waste, and underline the benefits of developing industrial demonstrators within the framework of international cooperation agreements. Waste retrieval capability will be simultaneously examined. Two types of waste have to be handled underground in Andra's repository. The HLW disposal package for vitrified waste is a 2 ton carbon steel cylindrical canister with a diameter of 600 mm. The weight of ILW-LL concrete disposal packages range from a minimum of 6 tonnes to over 20 tonnes, and their volume from approximately 5 to 10 m3. The underground transfer to the disposal drift requires moving the disposal package within a shielded transfer cask placed on a trailer. Transfer cask design has evolved since 2005, due to optimisation studies and as a result of industrial feedback from SKB. For HLW handling equipment two design options have been studied. In the first solution (Andra's Dossier 2005), the waste package are emplaced, one at a time, in the disposal drift by a pushing robot. Successive steps in design and proto-typing have lead to improve the design of the equipment and to gain confidence. Recently a fully integrated process has been successfully demonstrated, at full scale, (in a 100 m long mock up drift) as part of the EC funded ESDRED Project. This demonstrator is now on display in Andra's Technology Centre at Saudron, near the Bure Underground Laboratory. The second disposal option which has been investigated is based on a concept of utilising an external apparatus to push a row of

B-cell waste classification sampling plan

International Nuclear Information System (INIS)

HOBART, R.L.

1999-01-01

This report documents the methods used to collect samples and analyze data necessary to verify and/or determine the radionuclide content of the 324 Facility B-Cell decontamination and decommissioning waste stream

Rapid screening method for plutonium in mixed waste samples

International Nuclear Information System (INIS)

Somers, W.; Culp, T.; Miller, R.

1987-01-01

A waste stream sampling program was undertaken to determine those waste streams which contained hazardous constituents, and would therefore be regulated as a hazardous waste under the Resource Conservation and Recovery Act. The waste streams also had the potential of containing radioactive material, either plutonium, americium, or depleted uranium. Because of the potential for contamination with radioactive material, a method of rapidly screening the liquid samples for radioactive material was required. A counting technique was devised to count a small aliquot of a sample, determine plutonium concentration, and allow the sample to be shipped the same day they were collected. This technique utilized the low energy photons (x-rays) that accompany α decay. This direct, non-destructive x-ray analysis was applied to quantitatively determine Pu-239 concentrations in industrial samples. Samples contained a Pu-239, Am-241 mixture; the ratio and/or concentrations of these two radionuclides was not constant. A computer program was designed and implemented to calculate Pu-239 activity and concentration (g/ml) using the 59.5 keV Am-241 peak to determine Am-241's contribution to the 17 keV region. Am's contribution was subtracted, yielding net counts in the 17 keV region due to Pu. 2 figs., 1 tab

Waste tank characterization sampling limits

International Nuclear Information System (INIS)

Tusler, L.A.

1994-01-01

This document is a result of the Plant Implementation Team Investigation into delayed reporting of the exotherm in Tank 241-T-111 waste samples. The corrective actions identified are to have immediate notification of appropriate Tank Farm Operations Shift Management if analyses with potential safety impact exceed established levels. A procedure, WHC-IP-0842 Section 12.18, ''TWRS Approved Sampling and Data Analysis by Designated Laboratories'' (WHC 1994), has been established to require all tank waste sampling (including core, auger and supernate) and tank vapor samples be performed using this document. This document establishes levels for specified analysis that require notification of the appropriate shift manager. The following categories provide numerical values for analysis that may indicate that a tank is either outside the operating specification or should be evaluated for inclusion on a Watch List. The information given is intended to translate an operating limit such as heat load, expressed in Btu/hour, to an analysis related limit, in this case cesium-137 and strontium-90 concentrations. By using the values provided as safety flags, the analytical laboratory personnel can notify a shift manager that a tank is in potential violation of an operating limit or that a tank should be considered for inclusion on a Watch List. The shift manager can then take appropriate interim measures until a final determination is made by engineering personnel

Waste sludge resuspension and transfer: development program

International Nuclear Information System (INIS)

Weeren, H.O.; Mackey, T.S.

1980-02-01

The six Gunite waste tanks at Oak Ridge National Laboratory (ORNL) contain about 400,000 gal of sludge that has precipitated from solution and settled during the 35 years these tanks have been in service. Eventual decommissioning of the tanks has been proposed. The first part of this program is to resuspend the accumulated sludge, to transfer it to new storage tanks in Melton Valley, and to dispose of it by the shale-fracturing process. On the basis of preliminary information, a tentative operational concept was adopted. The sludge in each tank would be resuspended by hydraulic sluicing and pumped from the tank. This resuspended sludge would be treated as necessary to keep the particles in suspension and would be pumped to the new waste-storage tanks. Subsequently the sludge would be pumped from the tanks, combined with a cement-base mix, and disposed of by the shale-fracturing facility. Verification of the feasibility of this concept required development effort on characterization of the sludge and development of techniques for resuspending the sludge and for keeping it in suspension. These development efforts are described in this report. Sections of the report describe both the known properties of the sludge and the tests of grinding methods investigated, discuss tests of various suspenders, describe tests with cement-base mixes, summarize hot-cell tests on actual sludge samples, and describe tests that were made at a mockup of a Gunite tank installation. On the basis of the tests made, it was concluded that reslurrying and resuspension of the sludge is quite feasible and that the suspensions can be made compatible with cement mixes

Waste tank vapor project: Vapor space characterization of waste tank 241-BY-104: Results from samples collected on June 24, 1994

International Nuclear Information System (INIS)

Clauss, T.W.; Ligotke, M.W.; McVeety, B.D.; Pool, K.H.; Lucke, R.B.; Fruchter, J.S.; Goheen, S.C.

1994-11-01

This report describes results of the analyses of tank-headspace samples taken from Hanford waste Tank 241-BY-104 (referred to as Tank BY-104) on June 24, 1994. The Pacific Northwest Laboratory (PNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze inorganic and organic samples collected from the tank headspace. The sample job was designated S4019 and was performed by WHC on June 24, 1994 using the vapor sampling system (VSS). The results of the analyses are expected to be used in the determination of safety and toxicological issues related to the tank-headspace gas as described in the WHC report entitled Data Quality Objectives for Generic In-Tank Health and Safety Vapor Issue Resolution, WHC-SD-WM-DQO-002, Rev. 0. Sampling devices, including 16 sorbent trains (for inorganic analyses), and 5 SUMMA trademark canisters (for organic analyses), were supplied to the WHC sampling staff on June 20, 1994. Samples were taken (by WHC) on June 24. The samples were returned from the field on June 27. The inorganic samples delivered to PNL on chain-of-custody (COC) 006893 included 16 sorbent trains as described in Tables 2.2, 2.3, and 2.4. Additional inorganic blank spikes were obtained from related sample jobs. SUMMA trademark samples delivered to PNL on COC 006896 included one ambient air sample, one ambient-air sample through the sampling system, and three tank-headspace SUMMA trademark canister samples. The samples were inspected upon delivery to the 326/23B laboratory and logged into PNL laboratory record book 55408. Custody of the sorbent trains was transferred to PNL personnel performing the inorganic analysis and stored at refrigerated (≤10 degrees C) temperature until the time of analysis. Access to the 326/23B laboratory is limited to PNL personnel working on the waste-tank safety program

The Solid Waste Transfer Pit an FFTF [Fast Flux Test Facility] station for fuel and reflector transfer

International Nuclear Information System (INIS)

Milewski, A.J.; Thomson, J.D.

1985-01-01

This paper describes the design aspects of a proposed FFTF facility called the Solid Waste Transfer Pit (SWTP). This new facility will be used as a transfer station for reflectors and/or low decay heat fuel assemblies when movement from the Fuel Storage Facility for sodium removal and subsequent reprocessing is to be accomplished. The SWTP utilizes a shielded cell cover with an integral manually-operated gate valve for top loading of a transfer vessel. The cost-effective design discussed herein provides a practical approach using state-of-the-art concepts while assuring safe and reliable operation

Sampling and analyses of SRP high-level waste sludges

International Nuclear Information System (INIS)

Stone, J.A.; Kelley, J.A.; McMillan, T.S.

1976-08-01

Twelve 3-liter samples of high-heat waste sludges were collected from four Savannah River Plant waste tanks with a hydraulically operated sample collector of unique design. Ten of these samples were processed in Savannah River Laboratory shielded cell facilities, yielding 5.3 kg of washed, dried sludge products for waste solidification studies. After initial drying, each batch was washed by settling and decantation to remove the bulk of soluble salts and then was redried. Additional washes were by filtration, followed by final drying. Conclusions from analyses of samples taken during the processing steps were: (a) the raw sludges contained approximately 80 wt percent soluble salts, most of which were removed by the washes; (b) 90 Sr and 238 , 239 Pu remained in the sludges, but most of the 137 Cs was removed by washing; (c) small amounts of sodium, sulfate, and 137 Cs remained in the sludges after thorough washing; (d) no significant differences were found in sludge samples taken from different risers of one waste tank. Chemical and radiometric compositions of the sludge product from each tank were determined. The sludges had diverse compositions, but iron, manganese, aluminum, and uranium were principal elements in each sludge. 90 Sr was the predominant radionuclide in each sludge product

Application of Direct Assessment Approaches and Methodologies to Cathodically Protected Nuclear Waste Transfer Lines

International Nuclear Information System (INIS)

Dahl, Megan M.; Pikas, Joseph; Edgemon, Glenn L.; Philo, Sarah

2013-01-01

The U.S. Department of Energy's (DOE) Hanford Site is responsible for the safe storage, retrieval, treatment, and disposal of approximately 54 million gallons (204 million liters) of radioactive waste generated since the site's inception in 1943. Today, the major structures involved in waste management at Hanford include 149 carbon steel single-shell tanks, 28 carbon-steel double-shell tanks, plus a network of buried metallic transfer lines and ancillary systems (pits, vaults, catch tanks, etc.) required to store, retrieve, and transfer waste within the tank farm system. Many of the waste management systems at Hanford are still in use today. In response to uncertainties regarding the structural integrity of these systems,' an independent, comprehensive integrity assessment of the Hanford Site piping system was performed. It was found that regulators do not require the cathodically protected pipelines located within the Hanford Site to be assessed by External Corrosion Direct Assessment (ECDA) or any other method used to ensure integrity. However, a case study is presented discussing the application of the direct assessment process on pipelines in such a nuclear environment. Assessment methodology and assessment results are contained herein. An approach is described for the monitoring, integration of outside data, and analysis of this information in order to identify whether coating deterioration accompanied by external corrosion is a threat for these waste transfer lines

Radioactivity in waste water samples from COGEMA supplied by Greenpeace

International Nuclear Information System (INIS)

Reinen, H.A.J.M.; Kwakman, P.J.M.; Overwater, R.M.W.; Tax, R.B.; Nissan, L.A.

1999-01-01

The environmental organization Greenpeace sampled waste water from the reprocessing plant COGEMA in La Hague, France, in May 1999. On request of the Inspection Environmental Hygiene, The Dutch National Institute for Public Health and Environmental Protection (RIVM) determined the radioactivity of the waste water samples. 5 refs

Thermal modeling of core sampling in flammable gas waste tanks. Part 1: Push-mode sampling

International Nuclear Information System (INIS)

Unal, C.; Stroh, K.; Pasamehmetoglu, K.O.

1997-01-01

The radioactive waste stored in underground storage tanks at Hanford site is routinely being sampled for waste characterization purposes. The push- and rotary-mode core sampling is one of the sampling methods employed. The waste includes mixtures of sodium nitrate and sodium nitrite with organic compounds that can produce violent exothermic reactions if heated above 160 C during core sampling. A self-propagating waste reaction would produce very high temperatures that eventually result in failure of the tank and radioactive material releases to environment. A two-dimensional thermal model based on a lumped finite volume analysis method is developed. The enthalpy of each node is calculated from the first law of thermodynamics. A flash temperature and effective contact area concept were introduced to account the interface temperature rise. No maximum temperature rise exceeding the critical value of 60 C was found in the cases studied for normal operating conditions. Several accident conditions are also examined. In these cases it was found that the maximum drill bit temperature remained below the critical reaction temperature as long as a 30 scfm purge flow is provided the push-mode drill bit during sampling in rotary mode. The failure to provide purge flow resulted in exceeding the limiting temperatures in a relatively short time

Subsides for optimization of transfer of radioactive liquid waste from 99MO production plant to the waste treatment facility

International Nuclear Information System (INIS)

Rego, Maria Eugenia de Melo; Vicente, Roberto; Hiromoto, Goro

2013-01-01

The increasing need for radioisotopes lead Brazil to consider the domestic production of 99 Mo from fission of low enriched uranium targets. In order to meet the present demand of 99m Tc generators the planned 'end of irradiation' activity of 99 Mo is about 170 TBq per week. The radioactive waste from the production plant will be transferred to a waste treatment facility at the same site. The total activity of the actinides, fission and activation products present in the waste were predicted based on the fission yield and activation data for the irradiation conditions, such as composition and mass of uranium targets, irradiation time, neutron flux, production process and schedule, already established by the project management. The transfer of the waste from the production plant to the treatment facility will be done by means of special shielded packages. In the present study, the commercially available code Scale 6.0 was used to simulate the irradiation of the targets and the decay of radioactive products, assuming that an alkaline dissolution process would be performed on the targets before the removal and purification of 99 Mo. The assessment of the shielding required for the packages containing liquid waste was done using MicroShield 9 code. The results presented here are part of a project that aims at contributing to the design of the waste management system for the 99 Mo production facility. (author)

A Bayesian sampling strategy for hazardous waste site characterization

International Nuclear Information System (INIS)

Skalski, J.R.

1987-12-01

Prior knowledge based on historical records or physical evidence often suggests the existence of a hazardous waste site. Initial surveys may provide additional or even conflicting evidence of site contamination. This article presents a Bayes sampling strategy that allocates sampling at a site using this prior knowledge. This sampling strategy minimizes the environmental risks of missing chemical or radionuclide hot spots at a waste site. The environmental risk is shown to be proportional to the size of the undetected hot spot or inversely proportional to the probability of hot spot detection. 12 refs., 2 figs

Data quality objectives for the B-Cell waste stream classification sampling

International Nuclear Information System (INIS)

Barnett, J.M.

1998-01-01

This document defines the data quality objectives, (DQOS) for sampling the B-Cell racks waste stream. The sampling effort is concentrated on determining a ratio of Cs-137 to Sr-90 and Cs-137 to transuranics (TRU). Figure 1.0 shows the logic path of sampling effort. The flow chart begins with sample and data acquisition and progresses toward (a) statistical confidence and waste classification boundaries, (b) management decisions based on the input parameters and technical methods available, and (c) grout container volume/weight limits and radiation limits. The end result will be accurately classifying the B-Cell rack waste stream

High Level Waste Feed Delivery AZ-101 Batch Transfer to the Private Contractor Transfer and Mixing Process Improvements

International Nuclear Information System (INIS)

DUNCAN, G.P.

2000-01-01

The primary purpose of this business case is to provide Operations and Maintenance with a detailed transfer process review for the first High Level Waste (HLW) feed delivery to the Privatization Contractor (PC), AZ-101 batch transfer to PC. The Team was chartered to identify improvements that could be implemented in the field. A significant penalty can be invoked for not providing the quality, quantity, or timely delivery of HLW feed to the PC

Order of 13 December 1985 on the transfer to ENRESA of the Radioactive Waste Management Facility at Sierra Albarrana

International Nuclear Information System (INIS)

1984-01-01

This Order provides for the transfer of the Radioactive Waste Management Facility at Sierra Albarrana from the Junta de Energia Nuclear to ENRESA, the National Enterprise for Radioactive Waste; it also organises all stages of the transfer. (NEA) [fr

Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

Moujaes, S.F.; Lei, Y.M.

1994-01-01

Modeling free convection heat transfer in an cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rack. These waste containers are vertically emplaced in the borehole 300 meters below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3--4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions

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

Standard guide for sampling radioactive tank waste

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

1.1 This guide addresses techniques used to obtain grab samples from tanks containing high-level radioactive waste created during the reprocessing of spent nuclear fuels. Guidance on selecting appropriate sampling devices for waste covered by the Resource Conservation and Recovery Act (RCRA) is also provided by the United States Environmental Protection Agency (EPA) (1). Vapor sampling of the head-space is not included in this guide because it does not significantly affect slurry retrieval, pipeline transport, plugging, or mixing. 1.2 The values stated in inch-pound units are to be regarded as standard. No other units of measurement are included in this standard. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Subsides for optimization of transfer of radioactive liquid waste from {sup 99}MO production plant to the waste treatment facility

Energy Technology Data Exchange (ETDEWEB)

Rego, Maria Eugenia de Melo; Vicente, Roberto; Hiromoto, Goro, E-mail: maria.eugenia@ipen.br, E-mail: rvicente@ipen.br, E-mail: hiromoto@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

The increasing need for radioisotopes lead Brazil to consider the domestic production of {sup 99}Mo from fission of low enriched uranium targets. In order to meet the present demand of {sup 99m}Tc generators the planned 'end of irradiation' activity of {sup 99}Mo is about 170 TBq per week. The radioactive waste from the production plant will be transferred to a waste treatment facility at the same site. The total activity of the actinides, fission and activation products present in the waste were predicted based on the fission yield and activation data for the irradiation conditions, such as composition and mass of uranium targets, irradiation time, neutron flux, production process and schedule, already established by the project management. The transfer of the waste from the production plant to the treatment facility will be done by means of special shielded packages. In the present study, the commercially available code Scale 6.0 was used to simulate the irradiation of the targets and the decay of radioactive products, assuming that an alkaline dissolution process would be performed on the targets before the removal and purification of {sup 99}Mo. The assessment of the shielding required for the packages containing liquid waste was done using MicroShield 9 code. The results presented here are part of a project that aims at contributing to the design of the waste management system for the {sup 99}Mo production facility. (author)

Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

Moujaes, S.F.; Lei, Y.M.

1994-01-01

Modeling free convection heat transfer in a cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rock. These waste containers are vertically emplaced in the borehole 300 meters just below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3-4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions. The borehole wall temperature history has been found in the previous study, and was estimated to reach a maximum temperature of about 218 degrees C after 18 years from the emplacement. The temperature history of the rock surface is then used for the air-gap simulation. The problem includes convection and radiation heat transfer in a vertical enclosure. This paper will present the results of the convection in the air-gap over one thousand years after the containers' emplacement. During this long simulation period it was also observed that a multi-cellular air flow pattern can be generated in the air gap

DOUBLE-SHELL TANK WASTE TRANSFER LINE ENCASEMENT INTEGRITY ASSESSMENT TECHNOLOGY STUDY

International Nuclear Information System (INIS)

BOWER, R.R.

2006-01-01

The report provides various alternative methods of performing integrity assessment inspections of buried Hanford Double Shell Tank waste transfer line encasements, and provides method recommendations as an alternative to costly encasement pneumatic leak testing. A schedule for future encasement integrity assessments is also included

Development of on-site accident criteria for waste transfer casks

International Nuclear Information System (INIS)

Uldrich, E.D.

1989-01-01

Removal of radioactive waste must withstand the scrutiny of the public and various regulatory offices. Currently, there is no standard accident criteria or methodology for intra-site shipments at the Idaho National Engineering Laboratory (INEL). Since the radioactive waste transfer casks only carry material within the INEL site boundaries and are not used for normal over-the-road transport, the requirements of 10 CFR 71 Packaging and Transportation of Radioactive Material, do not provide suitable requirements for cask design or safety analyses. The objective is to develop realistically conservative accident scenarios consistent with the limited uses at the INEL for which the cask is approved

On the sample transport time of a pneumatic transfer system

International Nuclear Information System (INIS)

Kondo, Yoshihide

1983-01-01

The counts accumulated in measuring system are affected by the variations in transport time of the sample on cyclic activation experiments with a mechanical sample transfer system. In use of the pneumatic transfer system, which has been set up, the transport time is variable according to the differences as follows: The form, size and weight of samples, the pneumatic pressure and so on. Comprehending the relationships between the transpot time and these variable factors is essentially important to make experiments with this transfer system. (author)

Continued Evaluation of the Pulse-Echo Ultrasonic Instrument for Critical Velocity Determination during Hanford Tank Waste Transfer Operations - 12518

Energy Technology Data Exchange (ETDEWEB)

Denslow, Kayte M.; Bontha, Jagannadha R.; Adkins, Harold E.; Jenks, Jeromy W.J.; Burns, Carolyn A.; Schonewill, Philip P.; Hopkins, Derek F. [Pacific Northwest National Laboratory, Richland, Washington 99354 (United States); Thien, Michael G.; Wooley, Theodore A. [Washington River Protection Solutions, Richland, Washington 99354 (United States)

2012-07-01

The delivery of Hanford double-shell tank waste to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) will be governed by specific Waste Acceptance Criteria that are identified in ICD 19 - Interface Control Document for Waste Feed. Waste must be certified as acceptable before it can be delivered to the WTP. The fluid transfer velocity at which solid particulate deposition occurs in waste slurry transport piping (critical velocity) is a key waste parameter that must be accurately characterized to determine if the waste is acceptable for transfer to the WTP. In 2010 Washington River Protection Solutions and the Pacific Northwest National Laboratory began evaluating the ultrasonic PulseEcho instrument to accurately identify critical velocities in a horizontal slurry transport pipeline for slurries containing particles with a mean particle diameter of >50 micrometers. In 2011 the PulseEcho instrument was further evaluated to identify critical velocities for slurries containing fast-settling, high-density particles with a mean particle diameter of <15 micrometers. This two-year evaluation has demonstrated the ability of the ultrasonic PulseEcho instrument to detect the onset of critical velocity for a broad range of physical and rheological slurry properties that are likely encountered during the waste feed transfer operations between the Hanford tank farms and the WTP. (authors)

Savannah River Site sample and analysis plan for Clemson Technical Center waste

International Nuclear Information System (INIS)

Hagstrom, T.

1998-04-01

The purpose of this sampling and analysis plan is to determine the chemical, physical and radiological properties of the SRS radioactive Polychlorinated Biphenyl (PCB) liquid waste stream, to verify that it conforms to Waste Acceptance Criteria of the Department of Energy (DOE) East Tennessee Technology Park (ETTP) Toxic Substance Control Act (TSCA) Incineration Facility. Waste being sent to the ETTP TSCA Incinerator for treatment must be sufficiently characterized to ensure that the waste stream meets the waste acceptance criteria to ensure proper handling, classification, and processing of incoming waste to meet the Waste Storage and Treatment Facility's Operating Permits. This sampling and analysis plan is limited to WSRC container(s) of homogeneous or multiphasic radioactive PCB contaminated liquids generated in association with a treatability study at Clemson Technical Center (CTC) and currently stored at the WSRC Solid Waste Division Mixed Waste Storage Facility (MWSF)

Waste retrieval sluicing system vapor sampling and analysis plan for evaluation of organic emissions, process test phase III

International Nuclear Information System (INIS)

SASAKI, L.M.

1999-01-01

This sampling and analysis plan identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for vapor samples obtained to address vapor issues related to the sluicing of tank 241-C-106. Sampling will be performed in accordance with Waste Retrieval Sluicing System Emissions Collection Phase III (Jones 1999) and Process Test Plan Phase III, Waste Retrieval Sluicing System Emissions Collection (Powers 1999). Analytical requirements include those specified in Request for Ecology Concurrence on Draft Strategy/Path Forward to Address Concerns Regarding Organic Emissions from C-106 Sluicing Activities (Peterson 1998). The Waste Retrieval Sluicing System was installed to retrieve and transfer high-heat sludge from tank 241-C-106 to tank 241-AY-102, which is designed for high-heat waste storage. During initial sluicing of tank 241-C-106 in November 1998, operations were halted due to detection of unexpected high volatile organic compounds in emissions that exceeded regulatory permit limits. Several workers also reported smelling sharp odors and throat irritation. Vapor grab samples from the 296-C-006 ventilation system were taken as soon as possible after detection; the analyses indicated that volatile and semi-volatile organic compounds were present. In December 1998, a process test (phase I) was conducted in which the pumps in tanks 241-C-106 and 241-AY-102 were operated and vapor samples obtained to determine constituents that may be present during active sluicing of tank 241-C-106. The process test was suspended when a jumper leak was detected. On March 7, 1999, phase I1 of the process test was performed; the sluicing system was operated for approximately 7 hours and was ended using the controlled shutdown method when the allowable amount of solids were transferred to 241-AY-102. The phase II test was successful, however, further testing is required to obtain vapor samples at higher emission levels

Technology transfer on long-term radioactive waste management - a feasible option for small nuclear programmes?

International Nuclear Information System (INIS)

Mele, I.; Mathieson, J.

2007-01-01

The EU project CATT - Co-operation and technology transfer on long-term radioactive waste management for Member States with small nuclear programmes investigated the feasibility of countries with small nuclear programmes implementing long-term radioactive waste management solutions within their national borders, through collaboration on technology transfer with those countries with advanced disposal concepts. The main project objective was to analyse the existing capabilities of technology owning Member States and the corresponding requirements of potential technology acquiring Member States and, based on the findings, to develop a number of possible collaboration models and scenarios that could be used in a technology transfer scheme. The project CATT was performed as a specific support action under the EU sixth framework programme and it brought together waste management organisations from six EU Member States: UK, Bulgaria, Germany, Lithuania, Slovenia and Sweden. In addition, the EC Joint Research Centre from the Netherlands also participated as a full partner. The paper summarises the analyses performed and the results obtained within the project. (author)

Waste Retrieval Sluicing System Campaign Number 3 Solids Volume Transferred Calculation

International Nuclear Information System (INIS)

CAROTHERS, K.G.

1999-01-01

Waste Retrieval Sluicing System (WRSS) operations at tank 241-C-106 began on Wednesday, November 18, 1998. The purpose of this system is to retrieve and transfer the high-heat sludge from the tank for storage in double-shell tank 241-AY-102, thereby resolving the high-heat safety issue for the tank, and to demonstrate modernized past-practice retrieval technology for single-shell tank waste. Performance Agreement (PA) TWR 1.2.2, C-106 Sluicing, was established by the Department of Energy, Office of River Protection (ORP) for achieving completion of sluicing retrieval of waste from tank 241-C-106 by September 30, 1999. This level of sludge removal is defined in the PA as either removal of approximately 72 inches of sludge or removal of 172,000 gallons of sludge (approximately 62 inches) and less than 6,000 gallons (approximately 2 inches) of sludge removal per 12 hour sluice batch for three consecutive batches. Preliminary calculations of the volume of tank 241-C-106 sludge removed as of September 29, 1999 were provided to ORP documenting completion of PA TWR 1.2.2 (Allen 1999a). The purpose of this calculation is to document the final sludge volume removed from tank 241-C-106 up through September 30, 1999. Additionally, the results of an extra batch completed October 6, 1999 is included to show the total volume of sludge removed through the end of WRSS operations. The calculation of the sludge volume transferred from the tank is guided by engineering procedure HNF-SD-WM-PROC-021, Section 15.0,Rev. 3, sub-section 4.4, ''Calculation of Sludge Transferred.''

Waste Retrieval Sluicing System Campaign Number 3 Solids Volume Transferred Calculation

International Nuclear Information System (INIS)

CAROTHERS, K.G.

1999-01-01

Waste Retrieval Sluicing System (WRSS) operations at tank 241-C-106 began on Wednesday, November 18,1998. The purpose of this system is to retrieve and transfer the high-heat sludge from the tank for storage in double-shell tank 241-AY-102, thereby resolving the high-heat safety issue for the tank, and to demonstrate modernized past-practice retrieval technology for single-shell tank waste. Performance Agreement (PA) TWR 1.2.2, C-106 Sluicing, was established by the Department of Energy, Office of River Protection (ORP) for achieving completion of sluicing retrieval of waste from tank 241-C-106 by September 30,1999. This level of sludge removal is defined in the PA as either removal of approximately 72 inches of sludge or removal of 172,000 gallons of sludge (approximately 62 inches) and less than 6,000 gallons (approximately 2 inches) of sludge removal per 12 hour sluice batch for three consecutive batches. Preliminary calculations of the volume of tank 241-C-106 sludge removed as of September 29, 1999 were provided to ORP documenting completion of PA TWR 1.2.2 (Allen 1999a). The purpose of this calculation is to document the final sludge volume removed from tank 241-C-106 up through September 30, 1999. Additionally, the results of an extra batch completed October 6, 1999 is included to show the total volume of sludge removed through the end of WRSS operations. The calculation of the sludge volume transferred from the tank is guided by engineering procedure HNF-SD-WM-PROC-021, Section 15.0,Rev. 3, sub-section 4.4, ''Calculation of Sludge Transferred.''

Waste minimization in analytical chemistry through innovative sample preparation techniques

International Nuclear Information System (INIS)

Smith, L. L.

1998-01-01

Because toxic solvents and other hazardous materials are commonly used in analytical methods, characterization procedures result in significant and costly amount of waste. We are developing alternative analytical methods in the radiological and organic areas to reduce the volume or form of the hazardous waste produced during sample analysis. For the radiological area, we have examined high-pressure, closed-vessel microwave digestion as a way to minimize waste from sample preparation operations. Heated solutions of strong mineral acids can be avoided for sample digestion by using the microwave approach. Because reactivity increases with pressure, we examined the use of less hazardous solvents to leach selected contaminants from soil for subsequent analysis. We demonstrated the feasibility of this approach by extracting plutonium from a NET reference material using citric and tartaric acids with microwave digestion. Analytical results were comparable to traditional digestion methods, while hazardous waste was reduced by a factor often. We also evaluated the suitability of other natural acids, determined the extraction performance on a wider variety of soil types, and examined the extraction efficiency of other contaminants. For the organic area, we examined ways to minimize the wastes associated with the determination of polychlorinated biphenyls (PCBs) in environmental samples. Conventional methods for analyzing semivolatile organic compounds are labor intensive and require copious amounts of hazardous solvents. For soil and sediment samples, we have a method to analyze PCBs that is based on microscale extraction using benign solvents (e.g., water or hexane). The extraction is performed at elevated temperatures in stainless steel cells containing the sample and solvent. Gas chromatography-mass spectrometry (GC/MS) was used to quantitate the analytes in the isolated extract. More recently, we developed a method utilizing solid-phase microextraction (SPME) for natural

Transuranic waste characterization sampling and analysis methods manual. Revision 1

International Nuclear Information System (INIS)

Suermann, J.F.

1996-04-01

This Methods Manual provides a unified source of information on the sampling and analytical techniques that enable Department of Energy (DOE) facilities to comply with the requirements established in the current revision of the Transuranic Waste Characterization Quality Assurance Program Plan (QAPP) for the Waste Isolation Pilot Plant (WIPP) Transuranic (TRU) Waste Characterization Program (the Program) and the WIPP Waste Analysis Plan. This Methods Manual includes all of the testing, sampling, and analytical methodologies accepted by DOE for use in implementing the Program requirements specified in the QAPP and the WIPP Waste Analysis Plan. The procedures in this Methods Manual are comprehensive and detailed and are designed to provide the necessary guidance for the preparation of site-specific procedures. With some analytical methods, such as Gas Chromatography/Mass Spectrometry, the Methods Manual procedures may be used directly. With other methods, such as nondestructive characterization, the Methods Manual provides guidance rather than a step-by-step procedure. Sites must meet all of the specified quality control requirements of the applicable procedure. Each DOE site must document the details of the procedures it will use and demonstrate the efficacy of such procedures to the Manager, National TRU Program Waste Characterization, during Waste Characterization and Certification audits

A DOE manual: DOE Methods for Evaluating Environmental and Waste Management Samples

International Nuclear Information System (INIS)

Goheen, S.C.; McCulloch, M.; Riley, R.G.

1994-01-01

Waste Management inherently requires knowledge of the waste's chemical composition. The waste can often be analyzed by established methods; however, if the samples are radioactive, or are plagued by other complications, established methods may not be feasible. The US Department of Energy (DOE) has been faced with managing some waste types that are not amenable to standard or available methods, so new or modified sampling and analysis methods are required. These methods are incorporated into DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods), which is a guidance/methods document for sampling and analysis activities in support of DOE sites. It is a document generated by consensus of the DOE laboratory staff and is intended to fill the gap within existing guidance documents (e. g., the Environmental Protection Agency's (EPA's) Test Methods for Evaluating Solid Waste, SW-846), which apply to low-level or non-radioactive samples. DOE Methods fills the gap by including methods that take into account the complexities of DOE site matrices. The most recent update, distributed in October 1993, contained quality assurance (QA), quality control (QC), safety, sampling, organic analysis, inorganic analysis, and radioanalytical guidance as well as 29 methods. The next update, which will be distributed in April 1994, will contain 40 methods and will therefore have greater applicability. All new methods are either peer reviewed or labeled ''draft'' methods. Draft methods were added to speed the release of methods to field personnel

Influence of convective-energy transfer on calculated temperature distributions in proposed hard-rock nuclear waste repositories

Energy Technology Data Exchange (ETDEWEB)

Eaton, R R; Reda, D C [Sandia National Labs., Albuquerque, NM (USA)

1982-06-01

This study assesses the relative influence of convective-energy transfer on predicted temperature distributions for a nuclear-waste repository located in water-saturated rock. Using results for energy transfer by conduction only (no water motion) as a basis of comparison, it is shown that a considerable amount of energy can be removed from the repository by pumping out water that migrates into the drift from regions adjacent to the buried waste canisters. Furthermore, the results show that the influence of convective-energy transfer on mine drift cooling requirements can be significant for cases where the in-situ permeability of the rock is greater than one millidarcy (a regime potentially encountered in repository scenarios).

The potential for buoyant displacement gas release events in Tank 241-SY-102 after waste transfer from Tank 241-SY-101

International Nuclear Information System (INIS)

Wells, BE; Meyer, P.E.; Chen, G.

2000-01-01

Tank 241-SY-101 (SY-101) is a double-shell, radioactive waste storage tank with waste that, before the recent transfer and water back-dilution operations, was capable of retaining gas and producing buoyant displacement (BD) gas release events (GREs). Some BD GREs caused gas concentrations in the tank headspace to exceed the lower flammability limit (LFL). A BD GRE occurs when a portion of the nonconvective layer retains enough gas to become buoyant, rises to the waste surface, breaks up, and releases some of its stored gas. The installation of a mixer pump in 1993 successfully mitigated gas retention in the settled solids layer in SY-101 and has since prevented BD GREs. However, operation of the mixer pump over the years caused gas retention in the floating crust layer and a corresponding accelerated waste level growth. The accelerating crust growth trend observed in 1997--98 led to initiation of sequences of waste removal and water back-dilutions in December 1999. Waste is removed from the mixed slurry layer in Tank SY-101 and transferred into Tank 241-Sy-102 (SY-102). Water is then added back to dissolve soluble solids that retain gas. The initial transfer of 89,500 gallons of SY-101 waste, diluted in-line at 0.94:1 by volume with water, to SY-102 was conducted in December 1999. The second transfer of 230,000 gallons of original SY-101 waste, diluted approximately 0.9:1, was completed in January 2000, and the third transfer of 205,500 gallons of original SY-101 waste diluted at 0.9:1 was completed in March 2000

Sampling and analysis plan for sampling of liquid waste streams generated by 222-S Laboratory Complex operations

International Nuclear Information System (INIS)

Benally, A.B.

1997-01-01

This Sampling and Analysis Plan (SAP) establishes the requirements and guidelines to be used by the Waste Management Federal Services of Hanford, Inc. personnel in characterizing liquid waste generated at the 222-S Laboratory Complex. The characterization process to verify the accuracy of process knowledge used for designation and subsequent management of wastes consists of three steps: to prepare the technical rationale and the appendix in accordance with the steps outlined in this SAP; to implement the SAP by sampling and analyzing the requested waste streams; and to compile the report and evaluate the findings to the objectives of this SAP. This SAP applies to portions of the 222-S Laboratory Complex defined as Generator under the Resource Conservation and Recovery Act (RCRA). Any portion of the 222-S Laboratory Complex that is defined or permitted under RCRA as a treatment, storage, or disposal (TSD) facility is excluded from this document. This SAP applies to the liquid waste generated in the 222-S Laboratory Complex. Because the analytical data obtained will be used to manage waste properly, including waste compatibility and waste designation, this SAP will provide directions for obtaining and maintaining the information as required by WAC173-303

Low-level waste characterization plan for the WSCF Laboratory Complex

International Nuclear Information System (INIS)

Morrison, J.A.

1994-01-01

The Waste Characterization Plan for the Waste Sampling and Characterization Facility (WSCF) complex describes the organization and methodology for characterization of all waste streams that are transferred from the WSCF Laboratory Complex to the Hanford Site 200 Areas Storage and Disposal Facilities. Waste generated at the WSCF complex typically originates from analytical or radiological procedures. Process knowledge is derived from these operations and should be considered an accurate description of WSCF generated waste. Sample contribution is accounted for in the laboratory waste designation process and unused or excess samples are returned to the originator for disposal. The report describes procedures and processes common to all waste streams; individual waste streams; and radionuclide characterization methodology

Hazard evaluation for transfer of waste from tank 241-SY-101 to tank 241-SY-102

International Nuclear Information System (INIS)

Shultz, M.V.

1999-01-01

Tank 241-SY-101 waste level growth is an emergent, high priority issue. The purpose of this document is to record the hazards evaluation process and document potential hazardous conditions that could lead to the release of radiological and toxicological material from the proposed transfer of a limited quantity (approximately 100,000 gallons) of waste from Tank 241-SY-101 to Tank 241-SY-102. The results of the hazards evaluation were compared to the current Tank Waste Remediation System (TWRS) Basis for Interim Operation (HNF-SD-WM-BIO-001, 1998, Revision 1) to identify any hazardous conditions where Authorization Basis (AB) controls may not be sufficient or may not exist. Comparison to LA-UR-92-3196, A Safety Assessment for Proposed Pump Mixing Operations to Mitigate Episodic Gas Releases in Tank 241-SY-101, was also made in the case of transfer pump removal activities. Revision 1 of this document deletes hazardous conditions no longer applicable to the current waste transfer design and incorporates hazardous conditions related to the use of an above ground pump pit and overground transfer line. This document is not part of the AB and is not a vehicle for requesting authorization of the activity; it is only intended to provide information about the hazardous conditions associated with this activity. The AB Control Decision process will be used to determine the adequacy of controls and whether the proposed activity is within the AB. This hazard evaluation does not constitute an accident analysis

Mass transfer resistance in ASFF reactors for waste water treatment.

Science.gov (United States)

Ettouney, H M; Al-Haddad, A A; Abu-Irhayem, T M

1996-01-01

Analysis of mass transfer resistances was performed for an aerated submerged fixed-film reactor (ASFF) for the treatment of waste water containing a mixture of sucrose and ammonia. Both external and internal mass transfer resistances were considered in the analysis, and characterized as a function of feed flow-rate and concentration. Results show that, over a certain operating regime, external mass transfer resistance in the system was greater for sucrose removal than ammonia. This is because the reaction rates for carbon removal were much larger than those of nitrogen. As a result, existence of any form of mass transfer resistance caused by inadequate mixing or diffusion limitations, strongly affects the overall removal rates of carbon more than nitrogen. Effects of the internal måss transfer resistance were virtually non-existent for ammonia removal. This behaviour was found over two orders of magnitude range for the effective diffusivity for ammonia, and one order of magnitude for the film specific surface area. However, over the same parameters' range, it is found that sucrose removal was strongly affected upon lowering its effective diffusivity and increasing the film specific surface area.

Techniques for sampling nuclear waste tank contents and in situ measurement of activity

International Nuclear Information System (INIS)

Lawrence, R.C.

1978-04-01

A study was conducted to develop suitable sampling equipment and techniques for characterizing the mechanical properties of nuclear wastes; identifying effective means of measuring radiation levels, temperatures, and neutron fluxes in situ in wastes; and developing a waste core sampler. A portable, stainless steel probe was developed which is placed in the tank through a riser. This probe is built for the insertion of instrumentation that can measure the contents of the tank at any level and take temperature, radiation, and neutron activation readings with reliable accuracy. A simple and reliable instrument for the in situ extraction of waste materials ranging from liquid to concrete-like substances was also developed. This portable, stainless steel waste core sampler can remove up to one liter of radioactive waste from tanks for transportation to hot cell laboratories for analysis of hardness, chemical form, and isotopic content. A cask for transporting the waste samples from the tanks to the laboratory under radiation-protected conditions was also fabricated. This cask was designed with a ''boot'' or inner-seal liner to contain any radioactive wastes that might remain on the outside of the waste core sampling device

System Performance Testing of the Pulse-Echo Ultrasonic Instrument for Critical Velocity Determination during Hanford Tank Waste Transfer Operations - 13584

Energy Technology Data Exchange (ETDEWEB)

Denslow, Kayte M.; Bontha, Jagannadha R.; Adkins, Harold E.; Jenks, Jeromy W.J.; Hopkins, Derek F. [Pacific Northwest National Laboratory, Richland, Washington 99354 (United States); Thien, Michael G.; Kelly, Steven E.; Wooley, Theodore A. [Washington River Protection Solutions, Richland, Washington 99354 (United States)

2013-07-01

The delivery of Hanford double-shell tank waste to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is governed by specific Waste Acceptance Criteria that are identified in ICD 19 - Interface Control Document for Waste Feed. Waste must be certified as acceptable before it can be delivered to the WTP. The fluid transfer velocity at which solid particulate deposition occurs in waste slurry transport piping (critical velocity) is a key waste acceptance parameter that must be accurately characterized to determine if the waste is acceptable for transfer to the WTP. Washington River Protection Solutions and the Pacific Northwest National Laboratory have been evaluating the ultrasonic PulseEcho instrument since 2010 for its ability to detect particle settling and determine critical velocity in a horizontal slurry transport pipeline for slurries containing particles with a mean particle diameter of =14 micrometers (μm). In 2012 the PulseEcho instrument was further evaluated under WRPS' System Performance test campaign to identify critical velocities for slurries that are expected to be encountered during Hanford tank waste retrieval operations or bounding for tank waste feed. This three-year evaluation has demonstrated the ability of the ultrasonic PulseEcho instrument to detect the onset of critical velocity for a broad range of physical and rheological slurry properties that are likely encountered during the waste feed transfer operations between the Hanford tank farms and the WTP. (authors)

Environmental impacts of waste produced from processing of different uraniferous rock samples

Directory of Open Access Journals (Sweden)

Ibrahim E. El Aassy

2016-07-01

Full Text Available Radon exhalation rates from five studied laboratory waste samples resulted from five different sedimentary rock types named sandy dolostone, siltstone –two samples-, marly claystone and black shale were measured using ‘‘Sealed Can technique”. These rates were found to vary between 0.005 and 0.015 Bq m−2 h−1. A positive correlation was found between the radon exhalation rates and the radium activities. The emanation coefficients were calculated for these laboratory waste samples which varied between 0.0004 and 0.0007 according to the physical and chemical characterize of the wastes. These results are partially in accordance with autonite acid leached tailings on laboratory scale (USA. These results led us to pay attention about the effect and impact of these wastes on the environment.

Transferring knowledge about high-level waste repositories: An ethical consideration

International Nuclear Information System (INIS)

Berndes, S.; Kornwachs, K.

1996-01-01

The purpose of this paper is to present requirements to Information and Documentation Systems for high-level waste repositories from an ethical point of view. A structured synopsis of ethical arguments used by experts from Europe and America is presented. On the one hand the review suggests to reinforce the obligation to transfer knowledge about high level waste repositories. This obligation is reduced on the other hand by the objection that ethical obligations are dependent on the difference between our and future civilizations. This reflection results in proposing a list of well-balanced ethical arguments. Then a method is presented which shows how scenarios of possible future civilizations for different time horizons and related ethical arguments are used to justify requirements to the Information and Documentation System

Compatibility Grab Sampling and Analysis Plan for FY 2000

International Nuclear Information System (INIS)

SASAKI, L.M.

1999-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for grab samples obtained to address waste compatibility. It is written in accordance with requirements identified in Data Quality Objectives for Tank Farms Waste Compatibility Program (Mulkey et al. 1999) and Tank Farm Waste Transfer Compatibility Program (Fowler 1999). In addition to analyses to support Compatibility, the Waste Feed Delivery program has requested that tank samples obtained for Compatibility also be analyzed to confirm the high-level waste and/or low-activity waste envelope(s) for the tank waste (Baldwin 1999). The analytical requirements to confirm waste envelopes are identified in Data Quality Objectives for TWRS Privatization Phase I: Confirm Tank T is an Appropriate Feed Source for Low-Activity Waste Feed Batch X (Nguyen 1999a) and Data Quality Objectives for RPP Privatization Phase I: Confirm Tank T is an Appropriate Feed Source for High-Level Waste Feed Batch X (Nguyen 1999b)

Data Quality Objectives For Selecting Waste Samples For Bench-Scale Reformer Treatability Studies

International Nuclear Information System (INIS)

Banning, D.L.

2011-01-01

This document describes the data quality objectives to select archived samples located at the 222-S Laboratory for Bench-Scale Reforming testing. The type, quantity, and quality of the data required to select the samples for Fluid Bed Steam Reformer testing are discussed. In order to maximize the efficiency and minimize the time to treat Hanford tank waste in the Waste Treatment and Immobilization Plant, additional treatment processes may be required. One of the potential treatment processes is the fluidized bed steam reformer. A determination of the adequacy of the fluidized bed steam reformer process to treat Hanford tank waste is required. The initial step in determining the adequacy of the fluidized bed steam reformer process is to select archived waste samples from the 222-S Laboratory that will be used in a bench scale tests. Analyses of the selected samples will be required to confirm the samples meet the shipping requirements and for comparison to the bench scale reformer (BSR) test sample selection requirements.

Knowledge and technology transfer to improve the municipal solid waste management system of Durango City, Mexico.

Science.gov (United States)

Valencia-Vázquez, Roberto; Pérez-López, Maria E; Vicencio-de-la-Rosa, María G; Martínez-Prado, María A; Rubio-Hernández, Rubén

2014-09-01

As society evolves its welfare level increases, and as a consequence the amount of municipal solid waste increases, imposing great challenges to municipal authorities. In developed countries, municipalities have established integrated management schemes to handle, treat, and dispose of municipal solid waste in an economical and environmentally sound manner. Municipalities of developing and transition countries are not exempted from the challenges involving municipal solid waste handling, but their task is not easy to accomplish since they face budget deficits, lack of knowledge, and deficiencies in infrastructure and equipment. In the northern territory of Mexico, the municipality of Durango is facing the challenge of increased volumes of waste with a lack of adequate facilities and infrastructure. This article analyses the evolution of the municipal solid waste management of Durango city, which includes actions such as proper facilities construction, equipment acquisition, and the implementation of social programmes. The World Bank, offering courses to municipal managers on landfill operation and waste management, promoted the process of knowledge and technology transfer. Thereafter, municipal authorities attended regional and some international workshops on waste management. In addition they followed suggestions of international contractors and equipment dealers with the intention to improve the situation of the waste management of the city. After a 15-year period, transfer of knowledge and technology resulted in a modern municipal solid waste management system in Durango municipality. The actual system did not reach the standard levels of an integrated waste management system, nevertheless, a functional evaluation shows clear indications that municipality actions have put them on the right pathway. © The Author(s) 2014.

[DOE method for evaluating environmental and waste management samples: Revision 1, Addendum 1

Energy Technology Data Exchange (ETDEWEB)

Goheen, S.C.

1995-04-01

The US Dapartment of Energy`s (DOE`s) environmental and waste management (EM) sampling and analysis activities require that large numbers of samples be analyzed for materials characterization, environmental surveillance, and site-remediation programs. The present document, DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods), is a supplemental resource for analyzing many of these samples.

[DOE method for evaluating environmental and waste management samples: Revision 1, Addendum 1

International Nuclear Information System (INIS)

Goheen, S.C.

1995-04-01

The US Dapartment of Energy's (DOE's) environmental and waste management (EM) sampling and analysis activities require that large numbers of samples be analyzed for materials characterization, environmental surveillance, and site-remediation programs. The present document, DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods), is a supplemental resource for analyzing many of these samples

B-Cell waste classification sampling and analysis plan

International Nuclear Information System (INIS)

HOBART, R.L.

1999-01-01

This report documents the methods used to collect and analyze samples to obtain data necessary to verify and/or determine the radionuclide content of the 324 Facility B-Cell decontamination and decommissioning waste stream

Sampling and transport of paraffin waste form from CWDS of nuclear power plant

International Nuclear Information System (INIS)

Lee, J. M.; Hwang, J. H.; Kim, C. R.; Park, J. W.

2000-01-01

Sampling and transport of paraffin waste form from concentrated waste drying system (CWDS) of domestic nuclear power plant were performed to collect the leaching characteristic data for the disposal of radioactive waste. Transport was performed according to the national regulations and the internal rules of the nuclear power plant. The sample of paraffin waste form was classified as L type package according to the regulation and radiation exposure of operator was measured in the range of 6 to 12 mrem that was less than the estimated amount

Micro-PIXE evaluation of radioactive cesium transfer in contaminated soil samples

International Nuclear Information System (INIS)

Fujishiro, F.; Ishii, K.; Matsuyama, S.; Arai, H.; Ishizaki, A.; Osada, N.; Sugai, H.; Kusano, K.; Nozawa, Y.; Yamauchi, S.; Karahashi, M.; Oshikawa, S.; Kikuchi, K.; Koshio, S.; Watanabe, K.; Suzuki, Y.

2014-01-01

Highlights: • There are radioactively contaminated soils having a radioactive cesium transfer of 0.01. • Micro-PIXE analysis has revealed an existence of phosphorus in a contaminated soil. • Radioactive cesium captured by phosphorus compound would be due to radioactive transfer. -- Abstract: Micro-PIXE analysis has been performed on two soil samples with high cesium activity concentrations. These soil samples were contaminated by fallout from the accident at Fukushima Daiichi Nuclear Power Plant. One exhibits a radioactive cesium transfer of ∼0.01, and the other shows a radioactive cesium transfer of less than 0.001, even though both samples have high cesium activity concentrations exceeding 10,000 Bq/kg. X-ray spectra and elemental images of the soil samples revealed the presence of chlorine, which can react with cesium to produce an inorganic soluble compound, and phosphorus-containing cesium-capturable organic compounds

Heat-transfer aspects of Stirling power generation using incinerator waste energy

Energy Technology Data Exchange (ETDEWEB)

Hsu, S.T.; Lin, F.Y.; Chiou, J.S. [National Cheng Kung University, Tainan, Taiwan (China). Department of Mechanical Engineering

2003-01-01

The integration of a free-piston Stirling engine with linear alternator and an incinerator is able to effectively recover the waste energy and generate electrical power. In this study, a cycle-averaged heat transfer model is employed to investigate the performance of a free-piston Stirling engine installed on an incinerator. With the input of source and sink temperatures and other realistic heat transfer coefficients, the efficiency and the optimal power output are estimated, and the effect induced by internal and external irreversibilities is also evaluated. The proposed approach and modeling results presented in this study provide valuable information for engineers and designers to recover energy from small-scale incinerators. (author)

Waste Receiving and Processing Module 2A waste certification strategy

International Nuclear Information System (INIS)

LeClair, M.D.; Pottmeyer, J.A.; Hyre, R.A.

1994-01-01

This document addresses the certification of Mixed Low Level Waste (MLLW) that will be treated in the Waste Receiving and Processing Facility Module 2A (WRAP 2A) and is destined for disposal in the MLLW trench of the Low Level Burial Grounds (LLBG). The MLLW that will be treated in WRAP 2A contains land disposal restricted and radioactive constituents. Certification of the treated waste is dependent on numerous waste management activities conducted throughout the WRAP 2A operation. These activities range from waste treatability testing conducted prior to WRAP 2A waste acceptance to overchecking final waste form quality prior to transferring waste to disposal. This document addresses the high level strategies and methodologies for certifying the final waste form. Integration among all design and verification activities that support final waste form quality assurance is also discussed. The information generated from this effort may directly support other ongoing activities including the WRAP 2A Waste Characterization Study, WRAP 2A Waste Analysis Plan development, Sample Plan development, and the WRAP 2A Data Management System functional requirements definition

BWIP-RANDOM-SAMPLING, Random Sample Generation for Nuclear Waste Disposal

International Nuclear Information System (INIS)

Sagar, B.

1989-01-01

1 - Description of program or function: Random samples for different distribution types are generated. Distribution types as required for performance assessment modeling of geologic nuclear waste disposal are provided. These are: - Uniform, - Log-uniform (base 10 or natural), - Normal, - Lognormal (base 10 or natural), - Exponential, - Bernoulli, - User defined continuous distribution. 2 - Method of solution: A linear congruential generator is used for uniform random numbers. A set of functions is used to transform the uniform distribution to the other distributions. Stratified, rather than random, sampling can be chosen. Truncated limits can be specified on many distributions, whose usual definition has an infinite support. 3 - Restrictions on the complexity of the problem: Generation of correlated random variables is not included

Municipal solid waste composition: Sampling methodology, statistical analyses, and case study evaluation

DEFF Research Database (Denmark)

Edjabou, Vincent Maklawe Essonanawe; Jensen, Morten Bang; Götze, Ramona

2015-01-01

Sound waste management and optimisation of resource recovery require reliable data on solid waste generation and composition. In the absence of standardised and commonly accepted waste characterisation methodologies, various approaches have been reported in literature. This limits both...... comparability and applicability of the results. In this study, a waste sampling and sorting methodology for efficient and statistically robust characterisation of solid waste was introduced. The methodology was applied to residual waste collected from 1442 households distributed among 10 individual sub......-areas in three Danish municipalities (both single and multi-family house areas). In total 17 tonnes of waste were sorted into 10-50 waste fractions, organised according to a three-level (tiered approach) facilitating,comparison of the waste data between individual sub-areas with different fractionation (waste...

Laser-induced breakdown spectroscopy for the real-time analysis of mixed waste samples containing Sr

International Nuclear Information System (INIS)

Barefield, J.E. II; Koskelo, A.C.; Multari, R.A.; Cremers, D.A.; Gamble, T.K.; Han, C.Y.

1995-01-01

In this report, the use of Laser-induced breakdown spectroscopy to analyze mixed waste samples containing Sr is discussed. The mixed waste samples investigated include vitrified waste glass and contaminated soil. Compared to traditional analysis techniques, the laser-based method is fast (i.e., analysis times on the order of minutes) and essentially waste free since little or no sample preparation is required. Detection limits on the order of pmm Sr were determined. Detection limits obtained using a fiber optic cable to deliver laser pulses to soil samples containing Cr, Zr, Pb, Be, Cu, and Ni will also be discussed

Household hazardous waste data for the UK by direct sampling.

Science.gov (United States)

Slack, Rebecca J; Bonin, Michael; Gronow, Jan R; Van Santen, Anton; Voulvoulis, Nikolaos

2007-04-01

The amount of household hazardous waste (HHW) disposed of in the United Kingdom (UK) requires assessment. This paper describes a direct analysis study carried out in three areas in southeast England involving over 500 households. Each participating householder was provided with a special bin in which to place items corresponding to a list of HHW. The amount of waste collected was split into nine broad categories: batteries, home maintenance (DIY), vehicle upkeep, pesticides, pet care, pharmaceuticals, photographic chemicals, household cleaners, and printer cartridges. Over 1 T of waste was collected from the sample households over a 32-week period, which would correspond to an estimated 51,000 T if extrapolated to the UK population for the same period or over 7,000 T per month. Details of likely disposal routes adopted by householders were also sought, demonstrating the different pathways selected for different waste categories. Co-disposal with residual household waste dominated for waste batteries and veterinary medicines, hence avoiding classification as hazardous waste under new UK waste regulations. The information can be used to set a baseline for the management of HHW and provides information for an environmental risk assessment of the disposal of such wastes to landfill.

Municipal solid waste composition: Sampling methodology, statistical analyses, and case study evaluation

International Nuclear Information System (INIS)

Edjabou, Maklawe Essonanawe; Jensen, Morten Bang; Götze, Ramona; Pivnenko, Kostyantyn; Petersen, Claus; Scheutz, Charlotte; Astrup, Thomas Fruergaard

2015-01-01

Highlights: • Tiered approach to waste sorting ensures flexibility and facilitates comparison of solid waste composition data. • Food and miscellaneous wastes are the main fractions contributing to the residual household waste. • Separation of food packaging from food leftovers during sorting is not critical for determination of the solid waste composition. - Abstract: Sound waste management and optimisation of resource recovery require reliable data on solid waste generation and composition. In the absence of standardised and commonly accepted waste characterisation methodologies, various approaches have been reported in literature. This limits both comparability and applicability of the results. In this study, a waste sampling and sorting methodology for efficient and statistically robust characterisation of solid waste was introduced. The methodology was applied to residual waste collected from 1442 households distributed among 10 individual sub-areas in three Danish municipalities (both single and multi-family house areas). In total 17 tonnes of waste were sorted into 10–50 waste fractions, organised according to a three-level (tiered approach) facilitating comparison of the waste data between individual sub-areas with different fractionation (waste from one municipality was sorted at “Level III”, e.g. detailed, while the two others were sorted only at “Level I”). The results showed that residual household waste mainly contained food waste (42 ± 5%, mass per wet basis) and miscellaneous combustibles (18 ± 3%, mass per wet basis). The residual household waste generation rate in the study areas was 3–4 kg per person per week. Statistical analyses revealed that the waste composition was independent of variations in the waste generation rate. Both, waste composition and waste generation rates were statistically similar for each of the three municipalities. While the waste generation rates were similar for each of the two housing types (single

Municipal solid waste composition: Sampling methodology, statistical analyses, and case study evaluation

Energy Technology Data Exchange (ETDEWEB)

Edjabou, Maklawe Essonanawe, E-mail: vine@env.dtu.dk [Department of Environmental Engineering, Technical University of Denmark, 2800 Kgs. Lyngby (Denmark); Jensen, Morten Bang; Götze, Ramona; Pivnenko, Kostyantyn [Department of Environmental Engineering, Technical University of Denmark, 2800 Kgs. Lyngby (Denmark); Petersen, Claus [Econet AS, Omøgade 8, 2.sal, 2100 Copenhagen (Denmark); Scheutz, Charlotte; Astrup, Thomas Fruergaard [Department of Environmental Engineering, Technical University of Denmark, 2800 Kgs. Lyngby (Denmark)

2015-02-15

Highlights: • Tiered approach to waste sorting ensures flexibility and facilitates comparison of solid waste composition data. • Food and miscellaneous wastes are the main fractions contributing to the residual household waste. • Separation of food packaging from food leftovers during sorting is not critical for determination of the solid waste composition. - Abstract: Sound waste management and optimisation of resource recovery require reliable data on solid waste generation and composition. In the absence of standardised and commonly accepted waste characterisation methodologies, various approaches have been reported in literature. This limits both comparability and applicability of the results. In this study, a waste sampling and sorting methodology for efficient and statistically robust characterisation of solid waste was introduced. The methodology was applied to residual waste collected from 1442 households distributed among 10 individual sub-areas in three Danish municipalities (both single and multi-family house areas). In total 17 tonnes of waste were sorted into 10–50 waste fractions, organised according to a three-level (tiered approach) facilitating comparison of the waste data between individual sub-areas with different fractionation (waste from one municipality was sorted at “Level III”, e.g. detailed, while the two others were sorted only at “Level I”). The results showed that residual household waste mainly contained food waste (42 ± 5%, mass per wet basis) and miscellaneous combustibles (18 ± 3%, mass per wet basis). The residual household waste generation rate in the study areas was 3–4 kg per person per week. Statistical analyses revealed that the waste composition was independent of variations in the waste generation rate. Both, waste composition and waste generation rates were statistically similar for each of the three municipalities. While the waste generation rates were similar for each of the two housing types (single

Characterization of a low-level radioactive waste grout: Sampling and test results

International Nuclear Information System (INIS)

Martin, P.F.C.; Lokken, R.O.

1992-12-01

WHC manages and operates the grout treatment facility at Hanford as part of a DOE program to clean up wastes stored at federal nuclear production sites. PNL provides support to the grout disposal program through pilot-scale tests, performance assessments, and formulation verification activities. in 1988 and 1989, over one million gallons of a low-level radioactive liquid waste was processed through the facility to produce a grout waste that was then deposited in an underground vault. The liquid waste was phosphate/sulfate waste (PSW) generated in decontamination of the N Reactor. PNL sampled and tested the grout produced during the second half of the PSW campaign to support quality verification activities prior to grout vault closure. Samples of grout were obtained by inserting nested-tube samplers into the grout slurry in the vault. After the grout had cured, the inner tube of the sampler was removed and the grout samples extracted. Tests for compressive strength, sonic velocity, and leach testing were used to assess grout quality; results were compared to those from pilot-scale test grouts made with a simulated PSW. The grout produced during the second half of the PSW campaign exceeded compressive strength and leachability formulation criteria. The nested tube samplers were effective in collecting samples of grout although their use introduced greater variability into the compressive strength data

Sampling methods and non-destructive examination techniques for large radioactive waste packages

International Nuclear Information System (INIS)

Green, T.H.; Smith, D.L.; Burgoyne, K.E.; Maxwell, D.J.; Norris, G.H.; Billington, D.M.; Pipe, R.G.; Smith, J.E.; Inman, C.M.

1992-01-01

Progress is reported on work undertaken to evaluate quality checking methods for radioactive wastes. A sampling rig was designed, fabricated and used to develop techniques for the destructive sampling of cemented simulant waste using remotely operated equipment. An engineered system for the containment of cooling water was designed and manufactured and successfully demonstrated with the drum and coring equipment mounted in both vertical and horizontal orientations. The preferred in-cell orientation was found to be with the drum and coring machinery mounted in a horizontal position. Small powdered samples can be taken from cemented homogeneous waste cores using a hollow drill/vacuum section technique with the preferred subsampling technique being to discard the outer 10 mm layer to obtain a representative sample of the cement core. Cement blends can be dissolved using fusion techniques and the resulting solutions are stable to gelling for periods in excess of one year. Although hydrochloric acid and nitric acid are promising solvents for dissolution of cement blends, the resultant solutions tend to form silicic acid gels. An estimate of the beta-emitter content of cemented waste packages can be obtained by a combination of non-destructive and destructive techniques. The errors will probably be in excess of +/-60 % at the 95 % confidence level. Real-time X-ray video-imaging techniques have been used to analyse drums of uncompressed, hand-compressed, in-drum compacted and high-force compacted (i.e. supercompacted) simulant waste. The results have confirmed the applicability of this technique for NDT of low-level waste. 8 refs., 12 figs., 3 tabs

DOE methods for evaluating environmental and waste management samples

International Nuclear Information System (INIS)

Goheen, S.C.; McCulloch, M.; Thomas, B.L.; Riley, R.G.; Sklarew, D.S.; Mong, G.M.; Fadeff, S.K.

1994-04-01

DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods) is a resource intended to support sampling and analytical activities for the evaluation of environmental and waste management samples from U.S. Department of Energy (DOE) sites. DOE Methods is the result of extensive cooperation from all DOE analytical laboratories. All of these laboratories have contributed key information and provided technical reviews as well as significant moral support leading to the success of this document. DOE Methods is designed to encompass methods for collecting representative samples and for determining the radioisotope activity and organic and inorganic composition of a sample. These determinations will aid in defining the type and breadth of contamination and thus determine the extent of environmental restoration or waste management actions needed, as defined by the DOE, the U.S. Environmental Protection Agency, or others. The development of DOE Methods is supported by the Laboratory Management Division of the DOE. Methods are prepared for entry into DOE Methods as chapter editors, together with DOE and other participants in this program, identify analytical and sampling method needs. Unique methods or methods consolidated from similar procedures in the DOE Procedures Database are selected for potential inclusion in this document. Initial selection is based largely on DOE needs and procedure applicability and completeness. Methods appearing in this document are one of two types. open-quotes Draftclose quotes or open-quotes Verified.close quotes. open-quotes Draftclose quotes methods that have been reviewed internally and show potential for eventual verification are included in this document, but they have not been reviewed externally, and their precision and bias may not be known. open-quotes Verifiedclose quotes methods in DOE Methods have been reviewed by volunteers from various DOE sites and private corporations

Handling, assessment, transport and disposal of tritiated waste materials at JET

International Nuclear Information System (INIS)

Newbert, G.; Haigh, A.; Atkins, G.

1995-01-01

All types of JET radioactive wastes are received for disposal at the Waste Handling Facility (WHF) which features a waste sorting and sampling station, a glove box, a compactor, and packaging and transfer systems. The WHF is operated as a contamination control area with monitored tritium discharges. Two main types of tritium monitors used are liquid scintillation counters and ionization chambers, and samples of various components and materials have now been assessed for tritium. The results so far indicate a widespread of tritium levels from 2Bq/g for cold gas transfer lines to 200kBq/g for in-vessel tiles. General soft housekeeping waste is assessed by a sniffing technique which has a limit of detection corresponding to 120Bq/g. Investigation of improved methods of tritium measurement and of component detritiation was made to facilitate future waste disposal. 8 refs., 6 figs., 2 tabs

Filtration and Leach Testing for REDOX Sludge and S-Saltcake Actual Waste Sample Composites

Energy Technology Data Exchange (ETDEWEB)

Shimskey, Rick W.; Billing, Justin M.; Buck, Edgar C.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Geeting, John GH; Hallen, Richard T.; Jenson, Evan D.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Snow, Lanee A.; Swoboda, Robert G.

2009-02-20

A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan.( ) The test program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Under test plan TP-RPP-WTP-467, eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. Under this test plan, a waste-testing program was implemented that included: • Homogenizing the archive samples by group as defined in the test plan • Characterizing the homogenized sample groups • Performing parametric leaching testing on each group for compounds of interest • Performing bench-top filtration/leaching tests in the hot cell for each group to simulate filtration and leaching activities if they occurred in the UFP2 vessel of the WTP Pretreatment Facility. This report focuses on filtration/leaching tests performed on two of the eight waste composite samples and follow-on parametric tests to support aluminum leaching results from those tests.

Waste minimization and pollution prevention technology transfer : the Airlie House Projects

International Nuclear Information System (INIS)

Gatrone, R.; McHenry, J.; Myron, H.; Thout, J. R.

1998-01-01

The Airlie House Pollution Prevention Technology Transfer Projects were a series of pilot projects developed for the US Department of Energy with the intention of transferring pollution prevention technology to the private sector. The concept was to develop small technology transfer initiatives in partnership with the private sector. Argonne National Laboratory developed three projects: the microscale chemistry in education project, the microscale cost benefit study project, and the Bethel New Life recycling trainee project. The two microscale chemistry projects focused on introducing microscale chemistry technologies to secondary and college education. These programs were inexpensive to develop and received excellent evaluations from participants and regulators. The Bethel New Life recycling trainee project provided training for two participants who helped identify recycling and source reduction opportunities in Argonne National Laboratory's solid waste stream. The pilot projects demonstrated that technology transfer initiatives can be developed and implemented with a small budget and within a short period of time. The essential components of the pilot projects were identifying target technologies that were already available, identifying target audiences, and focusing on achieving a limited but defined objective

Radiochemical separation of actinides for their determination in environmental samples and waste products

Energy Technology Data Exchange (ETDEWEB)

Gleisberg, B [Nuclear Engineering and Analytics Rossendorf, Inc. (VKTA), Dresden (Germany)

1997-03-01

The determination of low level activities of actinides in environmental samples and waste products makes high demands on radiochemical separation methods. Artificial and natural actinides were analyzed in samples form the surrounding areas of NPP and of uranium mines, incorporation samples, solutions containing radioactive fuel, solutions and solids resutling from the process, and in wastes. The activities are measured by {alpha}-spectrometry and {gamma}-spectrometry. (DG)

DOE methods for evaluating environmental and waste management samples

Energy Technology Data Exchange (ETDEWEB)

Goheen, S.C.; McCulloch, M.; Thomas, B.L.; Riley, R.G.; Sklarew, D.S.; Mong, G.M.; Fadeff, S.K. [eds.

1994-10-01

DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods) is a resource intended to support sampling and analytical activities for the evaluation of environmental and waste management samples from U.S. Department of Energy (DOE) sites. DOE Methods is the result of extensive cooperation from all DOE analytical laboratories. All of these laboratories have contributed key information and provided technical reviews as well as significant moral support leading to the success of this document. DOE Methods is designed to encompass methods for collecting representative samples and for determining the radioisotope activity and organic and inorganic composition of a sample. These determinations will aid in defining the type and breadth of contamination and thus determine the extent of environmental restoration or waste management actions needed, as defined by the DOE, the U.S. Environmental Protection Agency, or others. The development of DOE Methods is supported by the Analytical Services Division of DOE. Unique methods or methods consolidated from similar procedures in the DOE Procedures Database are selected for potential inclusion in this document. Initial selection is based largely on DOE needs and procedure applicability and completeness. Methods appearing in this document are one of two types, {open_quotes}Draft{close_quotes} or {open_quotes}Verified{close_quotes}. {open_quotes}Draft{close_quotes} methods that have been reviewed internally and show potential for eventual verification are included in this document, but they have not been reviewed externally, and their precision and bias may not be known. {open_quotes}Verified{close_quotes} methods in DOE Methods have been reviewed by volunteers from various DOE sites and private corporations. These methods have delineated measures of precision and accuracy.

DOE methods for evaluating environmental and waste management samples

International Nuclear Information System (INIS)

Goheen, S.C.; McCulloch, M.; Thomas, B.L.; Riley, R.G.; Sklarew, D.S.; Mong, G.M.; Fadeff, S.K.

1994-10-01

DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods) is a resource intended to support sampling and analytical activities for the evaluation of environmental and waste management samples from U.S. Department of Energy (DOE) sites. DOE Methods is the result of extensive cooperation from all DOE analytical laboratories. All of these laboratories have contributed key information and provided technical reviews as well as significant moral support leading to the success of this document. DOE Methods is designed to encompass methods for collecting representative samples and for determining the radioisotope activity and organic and inorganic composition of a sample. These determinations will aid in defining the type and breadth of contamination and thus determine the extent of environmental restoration or waste management actions needed, as defined by the DOE, the U.S. Environmental Protection Agency, or others. The development of DOE Methods is supported by the Analytical Services Division of DOE. Unique methods or methods consolidated from similar procedures in the DOE Procedures Database are selected for potential inclusion in this document. Initial selection is based largely on DOE needs and procedure applicability and completeness. Methods appearing in this document are one of two types, open-quotes Draftclose quotes or open-quotes Verifiedclose quotes. open-quotes Draftclose quotes methods that have been reviewed internally and show potential for eventual verification are included in this document, but they have not been reviewed externally, and their precision and bias may not be known. open-quotes Verifiedclose quotes methods in DOE Methods have been reviewed by volunteers from various DOE sites and private corporations. These methods have delineated measures of precision and accuracy

DOE Low-Level Waste Management Program perspective on technology transfer: opportunities and challenges

International Nuclear Information System (INIS)

Large, D.E.

1982-01-01

The Department of Energy's Low-Level Waste Management Program (DOE LLWMP) perspective in regard to transfer of LLWMP technology to current and potential users in both the commercial and defense sectors is discussed. Past, present, and future opportunities and challenges for the whole nuclear waste management are indicated. Elements considered include: historical and evolutionary events and activities; the purpose of the Program and its inherent opportunities and challenges; achievements and expected accomplishments; supporters and interactors; packaging and delivering technology; implementing and serving potential users; determining and meeting users' needs; and identifying and responding to opportunities and challenges. The low-level waste management effort to improve shallow land burial technology began in FY 1977 and has expanded to include waste treatment and alternative disposal methods. Milestones have been established and are used as principal management control items. This technology, the Program Product, is described and is made available. This year, the Program has drafted criteria for inclusion in a DOE order for radioactive waste management operations at DOE sites

Effect of waste mica on transfer factors of 134Cs to spinach and lettuce

International Nuclear Information System (INIS)

Sreenivasa Chari, M.; Manjaiah, K.M.; Sachdev, P.; Sachdev, M.S.

2011-01-01

A greenhouse pot culture experiment was conducted to study the effect of graded levels of waste mica (0, 10, 20 and 40 g kg -1 ) on reducing the radiocesium uptake by spinach (Spinacia olerecea L) and lettuce (Lactuca sativa L.) grown in 134 Cs-contaminated (at 37 k Bq kg -1 soil) Inceptisols, Vertisols and Ultisols. The biomass yield, and potassium content and its uptake by crops have been significantly improved by waste mica application. The crops grown in Vertisols recorded higher biomass yield, and K content and its uptake as compared with Inceptisols and Ultisols. The average 134 Cs transfer factor values recorded were: 0.21, 0.17 and 0.26 at the first cutting, 0.15, 0.12 and 0.28 at the second cutting and 0.07, 0.05 and 0.23 at the third cutting from Inceptisols, Vertisols and Ultisols, respectively. Waste mica significantly suppressed radiocesium uptake, the effect being more pronounced at 40 g mica kg -1 soil. There exists an inverse relationship between the 134 Cs transfer factors with plant potassium content and also the K uptake by the crops. (authors)

Sampling and characterization of radioactive liquid wastes; Muestreo y caracterizacion de desechos liquidos radiactivos

Energy Technology Data Exchange (ETDEWEB)

Zepeda R, C.; Monroy G, F.; Reyes A, T.; Lizcano, D. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Cruz C, A. C., E-mail: carla.zepeda@inin.gob.mx [SEP, Instituto Tecnologico de Orizaba, Av. Oriente 9, Col. Emiliano Zapata, 94320 Orizaba, Veracruz (Mexico)

2017-09-15

To define the management of radioactive liquid wastes stored in 200 L drums, its isotope and physicochemical characterization is essential. An adequate sampling, that is, representative and homogeneous, is fundamental to obtain reliable analytical results, therefore, in this work, the use of a sampling mechanism that allows collecting homogenous aliquots, in a safe way and minimizing the generation of secondary waste is proposed. With this mechanism, 56 drums of radioactive liquid wastes were sampled, which were characterized by gamma spectrometry, liquid scintillation, and determined the following physicochemical properties: ph, conductivity, viscosity, density and chemical composition by gas chromatography. 67.86% of the radioactive liquid wastes contains H-3 and of these, 47.36% can be released unconditionally, since it presents activities lower than 100 Bq/g. 94% of the wastes are acidic and 48% have viscosities <50 MPa s. (Author)

Heat transfer analysis of the geologic disposal of spent fuel and high-level waste storage canisters

International Nuclear Information System (INIS)

Allen, G.K.

1980-08-01

Near-field temperatures resulting from the storage of high-level waste canisters and spent unreprocessed fuel assembly canisters in geologic formations were determined. Preliminary design of the repository was modeled for a heat transfer computer code, HEATING5, which used the Crank-Nicolson finite difference method to evaluate transient heat transfer. The heat transfer system was evaluated with several two- and three-dimensional models which transfer heat by a combination of conduction, natural convention, and radiation. Physical properties of the materials in the model were based upon experimental values for the various geologic formations. The effects of canister spacing, fuel age, and use of an overpack were studied for the analysis of the spent fuel canisters; salt, granite, and basalt were considered as the storage media for spent fuel canisters. The effects of canister diameter and use of an overpack were studied for the analysis of the high-level waste canisters; salt was considered as the only storage media for high-level waste canisters. Results of the studies on spent fuel assembly canisters showed that the canisters could be stored in salt formations with a maximum heat loading of 134 kw/acre without exceeding the temperature limits set for salt stability. The use of an overpack had little effect on the peak canister temperatures. When the total heat load per acre decreased, the peak temperatures reached in the geologic formations decreased; however, the time to reach the peak temperatures increased. Results of the studies on high-level waste canisters showed that an increased canister diameter will increase the canister interior temperatures considerably; at a constant areal heat loading, a 381 mm diameter canister reached almost a 50 0 C higher temperature than a 305 mm diameter canister. An overpacked canister caused almost a 30 0 C temperature rise in either case

Pump Jet Mixing and Pipeline Transfer Assessment for High-Activity Radioactive Wastes in Hanford Tank 241-AZ-102

Energy Technology Data Exchange (ETDEWEB)

Y Onishi; KP Recknagle; BE Wells

2000-08-09

The authors evaluated how well two 300-hp mixer pumps would mix solid and liquid radioactive wastes stored in Hanford double-shell Tank 241-AZ-102 (AZ-102) and confirmed the adequacy of a three-inch (7.6-cm) pipeline system to transfer the resulting mixed waste slurry to the AP Tank Farm and a planned waste treatment (vitrification) plant on the Hanford Site. Tank AZ-102 contains 854,000 gallons (3,230 m{sup 3}) of supernatant liquid and 95,000 gallons (360 m{sup 3}) of sludge made up of aging waste (or neutralized current acid waste). The study comprises three assessments: waste chemistry, pump jet mixing, and pipeline transfer. The waste chemical modeling assessment indicates that the sludge, consisting of the solids and interstitial solution, and the supernatant liquid are basically in an equilibrium condition. Thus, pump jet mixing would not cause much solids precipitation and dissolution, only 1.5% or less of the total AZ-102 sludge. The pump jet mixing modeling indicates that two 300-hp mixer pumps would mobilize up to about 23 ft (7.0 m) of the sludge nearest the pump but would not erode the waste within seven inches (0.18 m) of the tank bottom. This results in about half of the sludge being uniformly mixed in the tank and the other half being unmixed (not eroded) at the tank bottom.

Pump Jet Mixing and Pipeline Transfer Assessment for High-Activity Radioactive Wastes in Hanford Tank 241-AZ-102

International Nuclear Information System (INIS)

Onishi, Y.; Recknagle, K.P.; Wells, B.E.

2000-01-01

The authors evaluated how well two 300-hp mixer pumps would mix solid and liquid radioactive wastes stored in Hanford double-shell Tank 241-AZ-102 (AZ-102) and confirmed the adequacy of a three-inch (7.6-cm) pipeline system to transfer the resulting mixed waste slurry to the AP Tank Farm and a planned waste treatment (vitrification) plant on the Hanford Site. Tank AZ-102 contains 854,000 gallons (3,230 m 3 ) of supernatant liquid and 95,000 gallons (360 m 3 ) of sludge made up of aging waste (or neutralized current acid waste). The study comprises three assessments: waste chemistry, pump jet mixing, and pipeline transfer. The waste chemical modeling assessment indicates that the sludge, consisting of the solids and interstitial solution, and the supernatant liquid are basically in an equilibrium condition. Thus, pump jet mixing would not cause much solids precipitation and dissolution, only 1.5% or less of the total AZ-102 sludge. The pump jet mixing modeling indicates that two 300-hp mixer pumps would mobilize up to about 23 ft (7.0 m) of the sludge nearest the pump but would not erode the waste within seven inches (0.18 m) of the tank bottom. This results in about half of the sludge being uniformly mixed in the tank and the other half being unmixed (not eroded) at the tank bottom

A preliminary evaluation of comminution and sampling strategies for radioactive cemented waste

Energy Technology Data Exchange (ETDEWEB)

Bilodeau, M.; Lastra, R.; Bouzoubaa, N. [Natural Resources Canada, Ottawa, ON (Canada); Chapman, M. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2011-07-01

Lixiviation of Hg, U and Cs contaminants and micro-encapsulation of cemented radioactive waste (CRW) are the two main components of a CRW stabilization research project carried out at Natural Resources Canada in collaboration with Atomic Energy of Canada Limited. Unmolding CRW from the storage pail, its fragmentation into a size range suitable for both processes and the collection of a representative sample are three essential steps for providing optimal material conditions for the two studies. Separation of wires, metals and plastic incorporated into CRW samples is also required. A comminution and sampling strategy was developed to address all those needs. Dust emissions and other health and safety concerns were given full consideration. Surrogate cemented waste (SCW) was initially used for this comminution study where Cu was used as a substitute for U and Hg. SCW was characterized as a friable material through the measurement of the Bond work index of 7.7 kWh/t. A mineralogical investigation and the calibration of material heterogeneity parameters of the sampling error model showed that Cu, Hg and Cs are finely disseminated in the cement matrix. A sampling strategy was built from the model and successfully validated with radioactive waste. A larger than expected sampling error was observed with U due to the formation of large U solid phases, which were not observed with the Cu tracer. SCW samples were crushed and ground under different rock fragmentation mechanisms: compression (jaw and cone crushers, rod mill), impact (ball mill), attrition, high voltage disintegration and high pressure water (and liquid nitrogen) jetting. Cryogenic grinding was also tested with the attrition mill. Crushing and grinding technologies were assessed against criteria that were gathered from literature surveys, experiential know-how and discussion with the client and field experts. Water jetting and its liquid nitrogen variant were retained for pail cutting and waste unmolding while

A preliminary evaluation of comminution and sampling strategies for radioactive cemented waste

International Nuclear Information System (INIS)

Bilodeau, M.; Lastra, R.; Bouzoubaa, N.; Chapman, M.

2011-01-01

Lixiviation of Hg, U and Cs contaminants and micro-encapsulation of cemented radioactive waste (CRW) are the two main components of a CRW stabilization research project carried out at Natural Resources Canada in collaboration with Atomic Energy of Canada Limited. Unmolding CRW from the storage pail, its fragmentation into a size range suitable for both processes and the collection of a representative sample are three essential steps for providing optimal material conditions for the two studies. Separation of wires, metals and plastic incorporated into CRW samples is also required. A comminution and sampling strategy was developed to address all those needs. Dust emissions and other health and safety concerns were given full consideration. Surrogate cemented waste (SCW) was initially used for this comminution study where Cu was used as a substitute for U and Hg. SCW was characterized as a friable material through the measurement of the Bond work index of 7.7 kWh/t. A mineralogical investigation and the calibration of material heterogeneity parameters of the sampling error model showed that Cu, Hg and Cs are finely disseminated in the cement matrix. A sampling strategy was built from the model and successfully validated with radioactive waste. A larger than expected sampling error was observed with U due to the formation of large U solid phases, which were not observed with the Cu tracer. SCW samples were crushed and ground under different rock fragmentation mechanisms: compression (jaw and cone crushers, rod mill), impact (ball mill), attrition, high voltage disintegration and high pressure water (and liquid nitrogen) jetting. Cryogenic grinding was also tested with the attrition mill. Crushing and grinding technologies were assessed against criteria that were gathered from literature surveys, experiential know-how and discussion with the client and field experts. Water jetting and its liquid nitrogen variant were retained for pail cutting and waste unmolding while

Collection, transfer and transport of waste: accounting of greenhouse gases and global warming contribution

DEFF Research Database (Denmark)

Eisted, Rasmus; Larsen, Anna Warberg; Christensen, Thomas Højlund

2009-01-01

) emissions were quantified. The emission factors were assigned a global warming potential (GWP) and aggregated into global warming factors (GWFs), which express the potential contribution to global warming from collection, transport and transfer of 1 tonne of wet waste. Six examples involving collection...

Determination of elements in industrial waste sample and TENORM using XRF Technique in Nuclear Malaysia

International Nuclear Information System (INIS)

Paulus, W.; Sarimah Mahat; Meor Yusoff Meor Sulaiman

2011-01-01

Industrial waste such as aluminium dross and TENORM waste, oil sludge has been used as sample in this research. Determination of main elements by using X-Ray Fluorescence (XRF) in Material Technology Group, Malaysian Nuclear Agency. Results shows that main elements in these samples, aluminium and silicon, respectively. Thereby, this research shows that XRF can be considered as one of the techniques that can be used in waste characterization and furthermore, it can help researchers and engineer in the research related to waste treatment especially radioactive waste. (author)

Statistical sampling plan for the TRU waste assay facility

International Nuclear Information System (INIS)

Beauchamp, J.J.; Wright, T.; Schultz, F.J.; Haff, K.; Monroe, R.J.

1983-08-01

Due to limited space, there is a need to dispose appropriately of the Oak Ridge National Laboratory transuranic waste which is presently stored below ground in 55-gal (208-l) drums within weather-resistant structures. Waste containing less than 100 nCi/g transuranics can be removed from the present storage and be buried, while waste containing greater than 100 nCi/g transuranics must continue to be retrievably stored. To make the necessary measurements needed to determine the drums that can be buried, a transuranic Neutron Interrogation Assay System (NIAS) has been developed at Los Alamos National Laboratory and can make the needed measurements much faster than previous techniques which involved γ-ray spectroscopy. The previous techniques are reliable but time consuming. Therefore, a validation study has been planned to determine the ability of the NIAS to make adequate measurements. The validation of the NIAS will be based on a paired comparison of a sample of measurements made by the previous techniques and the NIAS. The purpose of this report is to describe the proposed sampling plan and the statistical analyses needed to validate the NIAS. 5 references, 4 figures, 5 tables

An optimization model for collection, haul, transfer, treatment and disposal of infectious medical waste: Application to a Greek region.

Science.gov (United States)

Mantzaras, Gerasimos; Voudrias, Evangelos A

2017-11-01

The objective of this work was to develop an optimization model to minimize the cost of a collection, haul, transfer, treatment and disposal system for infectious medical waste (IMW). The model calculates the optimum locations of the treatment facilities and transfer stations, their design capacities (t/d), the number and capacities of all waste collection, transport and transfer vehicles and their optimum transport path and the minimum IMW management system cost. Waste production nodes (hospitals, healthcare centers, peripheral health offices, private clinics and physicians in private practice) and their IMW production rates were specified and used as model inputs. The candidate locations of the treatment facilities, transfer stations and sanitary landfills were designated, using a GIS-based methodology. Specifically, Mapinfo software with exclusion criteria for non-appropriate areas was used for siting candidate locations for the construction of the treatment plant and calculating the distance and travel time of all possible vehicle routes. The objective function was a non-linear equation, which minimized the total collection, transport, treatment and disposal cost. Total cost comprised capital and operation costs for: (1) treatment plant, (2) waste transfer stations, (3) waste transport and transfer vehicles and (4) waste collection bins and hospital boxes. Binary variables were used to decide whether a treatment plant and/or a transfer station should be constructed and whether a collection route between two or more nodes should be followed. Microsoft excel software was used as installation platform of the optimization model. For the execution of the optimization routine, two completely different software were used and the results were compared, thus, resulting in higher reliability and validity of the results. The first software was Evolver, which is based on the use of genetic algorithms. The second one was Crystal Ball, which is based on Monte Carlo

Micro-PIXE evaluation of radioactive cesium transfer in contaminated soil samples

Science.gov (United States)

Fujishiro, F.; Ishii, K.; Matsuyama, S.; Arai, H.; Ishizaki, A.; Osada, N.; Sugai, H.; Kusano, K.; Nozawa, Y.; Yamauchi, S.; Karahashi, M.; Oshikawa, S.; Kikuchi, K.; Koshio, S.; Watanabe, K.; Suzuki, Y.

2014-01-01

Micro-PIXE analysis has been performed on two soil samples with high cesium activity concentrations. These soil samples were contaminated by fallout from the accident at Fukushima Daiichi Nuclear Power Plant. One exhibits a radioactive cesium transfer of ˜0.01, and the other shows a radioactive cesium transfer of less than 0.001, even though both samples have high cesium activity concentrations exceeding 10,000 Bq/kg. X-ray spectra and elemental images of the soil samples revealed the presence of chlorine, which can react with cesium to produce an inorganic soluble compound, and phosphorus-containing cesium-capturable organic compounds.

Sampling plan design and analysis for a low level radioactive waste disposal program

International Nuclear Information System (INIS)

Hassig, N.L.; Wanless, J.W.

1989-01-01

Low-level wastes that are candidates for BRC (below regulatory concern) disposal must be subjected to an extensive monitoring program to insure the wastes meet (potential) bulk property and contamination concentration BRC criteria for disposal. This paper addresses the statistical implications of using various methods to verify BRC criteria. While surface and volumetric monitoring each have their advantages and disadvantages, a dual, sequential monitoring process is the preferred choice from a statistical reliability perspective. With dual monitoring, measurements on the contamination are verifiable, and sufficient to allow for a complete characterization of the wastes. As these characterizations become more reliable and stable, something less than 100% sampling may be possible for release of wastes for BRC disposal. This paper provides a survey of the issues involved in the selection of a monitoring and sampling program for the disposal of BRC wastes

ENGINEERING SPECIALTY ASSESSMENT OF TANK WASTE COMPATIBILITY REPORTING

International Nuclear Information System (INIS)

KNIGHT, M.A.

2003-01-01

This Engineering Specialty Assessment was conducted to review the Tank Farm Waste Transfer Compatibility Program to assess whether the program meets the needs of accelerated retrieval and closure and waste feed delivery and to identify areas and methods for streamlining the program. The assessment was conducted in June 2003 and resulted in two findings and thirteen observations. The assessment results indicate that significant opportunities exist for streamlining the program by reducing the number of criteria requiring evaluation from 21 to 11, with only six of the criteria requiring evaluation for the majority of transfers. The assessment identified areas where existing criteria require strengthening to ensure that the risks of undesirable solids precipitation, from either waste mixing or waste transfer, are minimized. The assessment further identified opportunities for using existing engineering tools to simplify the calculations involved with preparation of waste compatibility assessments. The need to ensure that a revision to the waste compatibility program is prepared to align the program criteria with those that will be implemented with the DSA approval was also identified. Finally, the assessment identified that corrective actions are required to implement a tank-by-tank PCB inventory within the Best Basis Inventory and to ensure that sample data from external waste generators is entered into the TWINS database

Automation of Sample Transfer and Counting on Fast Neutron ActivationSystem

International Nuclear Information System (INIS)

Dewita; Budi-Santoso; Darsono

2000-01-01

The automation of sample transfer and counting were the transfer processof the sample to the activation and counting place which have been done byswitch (manually) previously, than being developed by automaticallyprogrammed logic instructions. The development was done by constructed theelectronics hardware and software for that communication. Transfer timemeasurement is on seconds and was done automatically with an error 1.6 ms.The counting and activation time were decided by the user on seconds andminutes, the execution error on minutes was 8.2 ms. This development systemwill be possible for measuring short half live elements and cyclic activationprocesses. (author)

Umbrella sampling of proton transfer in a creatine-water system

Science.gov (United States)

Ivchenko, Olga; Bachert, Peter; Imhof, Petra

2014-04-01

Proton transfer reactions are among the most common processes in chemistry and biology. Proton transfer between creatine and surrounding solvent water is underlying the chemical exchange saturation transfer used as a contrast in magnetic resonance imaging. The free energy barrier, determined by first-principles umbrella sampling simulations (EaDFT 3 kcal/mol) is in the same order of magnitude as the experimentally obtained activation energy. The underlying mechanism is a first proton transfer from the guanidinium group to the water pool, followed by a second transition where a proton is "transferred back" from the nearest water molecule to the deprotonated nitrogen atom of creatine.

Waste Sampling and Characterization Facility (WSCF)

International Nuclear Information System (INIS)

Bozich, J.L.

1993-07-01

This Maintenance Implementation Plan has been developed for maintenance functions associated with the Waste Sampling and Characterization Facility (WSCF). This plan is developed from the guidelines presented by Department of Energy (DOE) Order 4330.4A, Maintenance Management Program (DOE 1990), Chapter II. The objective of this plan is to provide baseline information for establishing and identifying WHC conformance programs and policies applicable to implementation of DOE order 4330.4A guidelines. In addition, this maintenance plan identifies the actions necessary to develop a cost-effective and efficient maintenance program at WSCF

Data Quality Objectives For Selecting Waste Samples To Test The Fluid Bed Steam Reformer Test

International Nuclear Information System (INIS)

Banning, D.L.

2010-01-01

This document describes the data quality objectives to select archived samples located at the 222-S Laboratory for Fluid Bed Steam Reformer testing. The type, quantity and quality of the data required to select the samples for Fluid Bed Steam Reformer testing are discussed. In order to maximize the efficiency and minimize the time to treat Hanford tank waste in the Waste Treatment and Immobilization Plant, additional treatment processes may be required. One of the potential treatment processes is the fluid bed steam reformer (FBSR). A determination of the adequacy of the FBSR process to treat Hanford tank waste is required. The initial step in determining the adequacy of the FBSR process is to select archived waste samples from the 222-S Laboratory that will be used to test the FBSR process. Analyses of the selected samples will be required to confirm the samples meet the testing criteria.

Heat Transfer Model of a Small-Scale Waste Glass Melter with Cold Cap Layer

Energy Technology Data Exchange (ETDEWEB)

Abboud, Alexander; Guillen, Donna Post; Pokorny, Richard

2016-09-01

At the Hanford site in the state of Washington, more than 56 million gallons of radioactive waste is stored in underground tanks. The cleanup plan for this waste is vitrification at the Waste Treatment Plant (WTP), currently under construction. At the WTP, the waste will be blended with glass-forming materials and heated to 1423K, then poured into stainless steel canisters to cool and solidify. A fundamental understanding of the glass batch melting process is needed to optimize the process to reduce cost and decrease the life cycle of the cleanup effort. The cold cap layer that floats on the surface of the glass melt is the primary reaction zone for the feed-to-glass conversion. The conversion reactions include water release, melting of salts, evolution of batch gases, dissolution of quartz and the formation of molten glass. Obtaining efficient heat transfer to this region is crucial to achieving high rates of glass conversion. Computational fluid dynamics (CFD) modeling is being used to understand the heat transfer dynamics of the system and provide insight to optimize the process. A CFD model was developed to simulate the DM1200, a pilot-scale melter that has been extensively tested by the Vitreous State Laboratory (VSL). Electrodes are built into the melter to provide Joule heating to the molten glass. To promote heat transfer from the molten glass into the reactive cold cap layer, bubbling of the molten glass is used to stimulate forced convection within the melt pool. A three-phase volume of fluid approach is utilized to model the system, wherein the molten glass and cold cap regions are modeled as separate liquid phases, and the bubbling gas and plenum regions are modeled as one lumped gas phase. The modeling of the entire system with a volume of fluid model allows for the prescription of physical properties on a per-phase basis. The molten glass phase and the gas phase physical properties are obtained from previous experimental work. Finding representative

Characterization, Leaching, and Filtration Testing for Tributyl Phosphate (TBP, Group 7) Actual Waste Sample Composites

Energy Technology Data Exchange (ETDEWEB)

Edwards, Matthew K.; Billing, Justin M.; Blanchard, David L.; Buck, Edgar C.; Casella, Amanda J.; Casella, Andrew M.; Crum, J. V.; Daniel, Richard C.; Draper, Kathryn E.; Fiskum, Sandra K.; Jagoda, Lynette K.; Jenson, Evan D.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Shimskey, Rick W.; Snow, Lanee A.; Swoboda, Robert G.

2009-03-09

.A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. The actual waste-testing program included homogenizing the samples by group, characterizing the solids and aqueous phases, and performing parametric leaching tests. The tributyl phosphate sludge (TBP, Group 7) is the subject of this report. The Group 7 waste was anticipated to be high in phosphorus as well as aluminum in the form of gibbsite. Both are believed to exist in sufficient quantities in the Group 7 waste to address leaching behavior. Thus, the focus of the Group 7 testing was on the removal of both P and Al. The waste-type definition, archived sample conditions, homogenization activities, characterization (physical, chemical, radioisotope, and crystal habit), and caustic leaching behavior as functions of time, temperature, and hydroxide concentration are discussed in this report. Testing was conducted according to TP-RPP-WTP-467.

Analytical characterization of high-level mixed wastes using multiple sample preparation treatments

International Nuclear Information System (INIS)

King, A.G.; Baldwin, D.L.; Urie, M.W.; McKinley, S.G.

1994-01-01

The Analytical Chemistry Laboratory at the Pacific Northwest Laboratory in Richland, Washington, is actively involved in performing analytical characterization of high-level mixed waste from Hanford's single shell and double shell tank characterization programs. A full suite of analyses is typically performed on homogenized tank core samples. These analytical techniques include inductively-coupled plasma-atomic emission spectroscopy, total organic carbon methods and radiochemistry methods, as well as many others, all requiring some type of remote sample-preparation treatment to solubilize the tank sludge material for analysis. Most of these analytical methods typically use a single sample-preparation treatment, inherently providing elemental information only. To better understand and interpret tank chemistry and assist in identifying chemical compounds, selected analytical methods are performed using multiple sample-preparation treatments. The sample preparation treatments used at Pacific Northwest Laboratory for this work with high-level mixed waste include caustic fusion, acid digestion, and water leach. The type of information available by comparing results from different sample-prep treatments includes evidence for the presence of refractory compounds, acid-soluble compounds, or water-soluble compounds. Problems unique to the analysis of Hanford tank wastes are discussed. Selected results from the Hanford single shell ferrocyanide tank, 241-C-109, are presented, and the resulting conclusions are discussed

Vapor space characterization of waste tank 241-C-101: Results from samples collected on 9/1/94

International Nuclear Information System (INIS)

Lucke, R.B.; Clauss, T.W.; Ligotke, M.W.

1995-11-01

This report describes results of the analyses of tank-headspace samples taken from the Hanford waste Tank 241-C-101 (referred to as Tank C-101) and the ambient air collected - 30 ft upwind near the tank and through the VSS near the tank. Pacific Northwest Laboratory (PNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and to analyze inorganic and organic analytes collected from the tank headspace and ambient air near the tank. The sample job was designated S4056, and samples were collected by WHC on September 1, 1994, using the vapor sampling system (VSS). The samples were inspected upon delivery to the 326/23B laboratory and logged into PNL record book 55408 before implementation of PNL Technical Procedure PNL-TVP-07. Custody of the sorbent traps was transferred to PNL personnel performing the inorganic analysis and stored at refrigerated (≤ 10 degrees C) temperature until the time of analysis. The canisters were stored in the 326/23B laboratory at ambient (25 degrees C) temperature until the time of the analysis. Access to the 326/23B laboratory is limited to PNL personnel working on the waste-tank safety program. Analyses described in this report were performed at PNL in the 300 area of the Hanford Reservation. Analytical methods that were used are described in the text. In summary, sorbent traps for inorganic analyses containing sample materials were either weighed (for water analysis) or desorbed with the appropriate aqueous solutions (for NH 3 , NO 2 , and NO analyses). The aqueous extracts were analyzed either by selective electrode or by ion chromatography (IC). Organic analyses were performed using cryogenic preconcentration followed by gas chromatography/mass spectrometry (GC/MS)

The radioactive waste management at IAEA laboratories

International Nuclear Information System (INIS)

Deron, S.; Ouvrard, R.; Hartmann, R.; Klose, H.

1992-10-01

The report gives a brief description of the nature of the radioactive wastes generated at the IAEA Laboratories in Seibersdorf, their origin and composition, their management and monitoring. The management of the radioactive waste produced at IAEA Laboratories in Seibersdorf is governed by the Technical Agreements of 1985 between the IAEA and the Austrian Health Ministry. In the period of 1982 to 1991 waste containers of low activity and radiotoxicity generated at laboratories other than the Safeguards Analytical Laboratory (SAL) were transferred to the FZS waste treatment and storage plant: The total activity contained in these drums amounted to < 65 MBq alpha activity; < 1030 MBq beta activity; < 2900 MBq gamma activity. The radioactive waste generated at SAL and transferred to the FZs during the same period included. Uranium contaminated solid burnable waste in 200 1 drums, uranium contaminated solid unburnable waste in 200 1 drums, uranium contaminated liquid unburnable waste in 30 1 bottles, plutonium contaminated solid unburnable waste in 200 1 drums. In the same period SAL received a total of 146 Kg uranium and 812 g plutonium and exported out of Austria, unused residues of samples. The balance, i.e.: uranium 39 kg, plutonium 133 g constitutes the increase of the inventory of reference materials, and unused residues awaiting export, accumulated at SAL and SIL fissile store as a result of SAL operation during this 10 year period. The IAEA reexports all unused residues of samples of radioactive and fissile materials analyzed at his laboratories, so that the amount of radioactive materials ending in the wastes treated and stored at FZS is kept to a minimum. 5 refs, 7 figs, 3 tabs

Waste Sampling and Characterization Facility (WSCF) Complex Safety Analysis

International Nuclear Information System (INIS)

MELOY, R.T.

2003-01-01

The Waste Sampling and Characterization Facility (WSCF) is an analytical laboratory complex on the Hanford Site that was constructed to perform chemical and low-level radiological analyses on a variety of sample media in support of Hanford Site customer needs. The complex is located in the 600 area of the Hanford Site, east of the 200 West Area. Customers include effluent treatment facilities, waste disposal and storage facilities, and remediation projects. Customers primarily need analysis results for process control and to comply with federal, Washington State, and US. Department of Energy (DOE) environmental or industrial hygiene requirements. This document was prepared to analyze the facility for safety consequences and includes the following steps: Determine radionuclide and highly hazardous chemical inventories; Compare these inventories to the appropriate regulatory limits; Document the compliance status with respect to these limits; and Identify the administrative controls necessary to maintain this status

Waste analysis plan for T Plant Complex

International Nuclear Information System (INIS)

Williams, J.F.

1996-01-01

Washington Administration Code 173-303-300 requires that a waste analysis plan (WAP) be provided by a treatment, storage, and/or disposal (TSD) unit to confirm their knowledge about a dangerous and/or mixed waste to ensure that the waste is managed properly. The specific objectives of the WAP are as follows: Ensure safe management of waste during treatment and storage; Ensure that waste generated during operational activities is properly designated in accordance with regulatory requirements; Provide chemical and physical analysis of representative samples of the waste stored for characterization and/or verification before the waste is transferred to another TSD unit; Ensure compliance with land disposal restriction (LDR) requirements for treated waste; and Provide basis for work plans that describes waste analysis for development of new treatment technologies

60-Day waste compatibility safety issues and final results for AY-102 grab samples

Energy Technology Data Exchange (ETDEWEB)

Nuzum, J.L.

1997-01-31

Four grab samples (2AY-96-15, 2AY-96-16, 2AY-96-17, and 2AY-96-18) were taken from Riser 15D of Tank 241-AY-102 on October 8, 1996, and received by 222-S Laboratory on October 8, 1996. These samples were analyzed in accordance with Compatibility Grab Sampling and Analysis Plan (TSAP) and Data Quality Objectives for Tank Farms Waste Compatibility Program (DQO) in support of the Waste Compatibility Program. No notifications were required based on sample results.

CANISTER TRANSFER SYSTEM DESCRIPTION DOCUMENT

International Nuclear Information System (INIS)

B. Gorpani

2000-01-01

The Canister Transfer System receives transportation casks containing large and small disposable canisters, unloads the canisters from the casks, stores the canisters as required, loads them into disposal containers (DCs), and prepares the empty casks for re-shipment. Cask unloading begins with cask inspection, sampling, and lid bolt removal operations. The cask lids are removed and the canisters are unloaded. Small canisters are loaded directly into a DC, or are stored until enough canisters are available to fill a DC. Large canisters are loaded directly into a DC. Transportation casks and related components are decontaminated as required, and empty casks are prepared for re-shipment. One independent, remotely operated canister transfer line is provided in the Waste Handling Building System. The canister transfer line consists of a Cask Transport System, Cask Preparation System, Canister Handling System, Disposal Container Transport System, an off-normal canister handling cell with a transfer tunnel connecting the two cells, and Control and Tracking System. The Canister Transfer System operating sequence begins with moving transportation casks to the cask preparation area with the Cask Transport System. The Cask Preparation System prepares the cask for unloading and consists of cask preparation manipulator, cask inspection and sampling equipment, and decontamination equipment. The Canister Handling System unloads the canister(s) and places them into a DC. Handling equipment consists of a bridge crane hoist,; DC--loading manipulator, lifting fixtures, and small canister staging racks. Once the cask has been unloaded, the Cask Preparation System decontaminates the cask exterior and returns it to the Carrier/Cask Handling System via the Cask Transport System. After the; DC--is fully loaded, the Disposal Container Transport System moves the; DC--to the Disposal Container Handling System for welding. To handle off-normal canisters, a separate off-normal canister

Canister Transfer System Description Document

International Nuclear Information System (INIS)

2000-01-01

The Canister Transfer System receives transportation casks containing large and small disposable canisters, unloads the canisters from the casks, stores the canisters as required, loads them into disposal containers (DCs), and prepares the empty casks for re-shipment. Cask unloading begins with cask inspection, sampling, and lid bolt removal operations. The cask lids are removed and the canisters are unloaded. Small canisters are loaded directly into a DC, or are stored until enough canisters are available to fill a DC. Large canisters are loaded directly into a DC. Transportation casks and related components are decontaminated as required, and empty casks are prepared for re-shipment. One independent, remotely operated canister transfer line is provided in the Waste Handling Building System. The canister transfer line consists of a Cask Transport System, Cask Preparation System, Canister Handling System, Disposal Container Transport System, an off-normal canister handling cell with a transfer tunnel connecting the two cells, and Control and Tracking System. The Canister Transfer System operating sequence begins with moving transportation casks to the cask preparation area with the Cask Transport System. The Cask Preparation System prepares the cask for unloading and consists of cask preparation manipulator, cask inspection and sampling equipment, and decontamination equipment. The Canister Handling System unloads the canister(s) and places them into a DC. Handling equipment consists of a bridge crane/hoist, DC loading manipulator, lifting fixtures, and small canister staging racks. Once the cask has been unloaded, the Cask Preparation System decontaminates the cask exterior and returns it to the Carrier/Cask Handling System via the Cask Transport System. After the DC is fully loaded, the Disposal Container Transport System moves the DC to the Disposal Container Handling System for welding. To handle off-normal canisters, a separate off-normal canister handling

M3FT-17OR0301070211 - Preparation of Hot Isostatically Pressed AgZ Waste Form Samples

Energy Technology Data Exchange (ETDEWEB)

Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jordan, Jacob A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-10-01

The production of radioactive iodine-bearing waste forms that exhibit long-term stability and are suitable for permanent geologic disposal has been the subject of substantial research interest. One potential method of iodine waste form production is hot isostatic pressing (HIP). Recent studies at Oak Ridge National Laboratory (ORNL) have investigated the conversion of iodine-loaded silver mordenite (I-AgZ) directly to a waste form by HIP. ORNL has performed HIP with a variety of sample compositions and pressing conditions. The base mineral has varied among AgZ (in pure and engineered forms), silver-exchanged faujasite, and silverexchanged zeolite A. Two iodine loading methods, occlusion and chemisorption, have been explored. Additionally, the effects of variations in temperature and pressure of the process have been examined, with temperature ranges of 525°C–1,100°C and pressure ranges of 100–300 MPa. All of these samples remain available to collaborators upon request. The sample preparation detailed in this document is an extension of that work. In addition to previously prepared samples, this report documents the preparation of additional samples to support stability testing. These samples include chemisorbed I-AgZ and pure AgI. Following sample preparation, each sample was processed by HIP by American Isostatic Presses Inc. and returned to ORNL for storage. ORNL will store the samples until they are requested by collaborators for durability testing. The sample set reported here will support waste form durability testing across the national laboratories and will provide insight into the effects of varied iodine content on iodine retention by the produced waste form and on potential improvements in waste form durability provided by the zeolite matrix.

Data Quality Objectives for Tank Farms Waste Compatibility Program

International Nuclear Information System (INIS)

BANNING, D.L.

1999-01-01

There are 177 waste storage tanks containing over 210,000 m 3 (55 million gal) of mixed waste at the Hanford Site. The River Protection Project (RPP) has adopted the data quality objective (DQO) process used by the U.S. Environmental Protection Agency (EPA) (EPA 1994a) and implemented by RPP internal procedure (Banning 1999a) to identify the information and data needed to address safety issues. This DQO document is based on several documents that provide the technical basis for inputs and decision/action levels used to develop the decision rules that evaluate the transfer of wastes. A number of these documents are presently in the process of being revised. This document will need to be revised if there are changes to the technical criteria in these supporting documents. This DQO process supports various documents, such as sampling and analysis plans and double-shell tank (DST) waste analysis plans. This document identifies the type, quality, and quantity of data needed to determine whether transfer of supernatant can be performed safely. The requirements in this document are designed to prevent the mixing of incompatible waste as defined in Washington Administrative Code (WAC) 173-303-040. Waste transfers which meet the requirements contained in this document and the Double-Shell Tank Waste Analysis Plan (Mulkey 1998) are considered to be compatible, and prevent the mixing of incompatible waste

Filtration and Leach Testing for PUREX Cladding Sludge and REDOX Cladding Sludge Actual Waste Sample Composites

Energy Technology Data Exchange (ETDEWEB)

Shimskey, Rick W.; Billing, Justin M.; Buck, Edgar C.; Casella, Amanda J.; Crum, Jarrod V.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Hallen, Richard T.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Swoboda, Robert G.

2009-03-02

A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan (Barnes and Voke 2006). The test program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Hanford Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Under test plan TP RPP WTP 467 (Fiskum et al. 2007), eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. Under this test plan, a waste testing program was implemented that included: • Homogenizing the archive samples by group as defined in the test plan. • Characterizing the homogenized sample groups. • Performing parametric leaching testing on each group for compounds of interest. • Performing bench-top filtration/leaching tests in the hot cell for each group to simulate filtration and leaching activities if they occurred in the UFP2 vessel of the WTP Pretreatment Facility. This report focuses on a filtration/leaching test performed using two of the eight waste composite samples. The sample groups examined in this report were the plutonium-uranium extraction (PUREX) cladding waste sludge (Group 3, or CWP) and reduction-oxidation (REDOX) cladding waste sludge (Group 4, or CWR). Both the Group 3 and 4 waste composites were anticipated to be high in gibbsite, thus requiring caustic leaching. WTP RPT 167 (Snow et al. 2008) describes the homogenization, characterization, and parametric leaching activities before benchtop filtration/leaching testing of these two waste groups. Characterization and initial parametric data in that report were used to plan a single filtration/leaching test using a blend of both wastes. The test focused on filtration testing of the waste and caustic leaching for aluminum, in the form

Characterization of the March 2017 Tank 15 Waste Removal Slurry Sample (Combination of Slurry Samples HTF-15-17-28 and HTF-15-17-29)

Energy Technology Data Exchange (ETDEWEB)

Reboul, S. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); King, W. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Coleman, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-05-09

Two March 2017 Tank 15 slurry samples (HTF-15-17-28 and HTF-15-17-29) were collected during the second bulk waste removal campaign and submitted to SRNL for characterization. At SRNL, the two samples were combined and then characterized by a series of physical, elemental, radiological, and ionic analysis methods. Sludge settling as a function of time was also quantified. The characterization results reported in this document are consistent with expectations based upon waste type, process knowledge, comparisons between alternate analysis techniques, and comparisons with the characterization results obtained for the November 2016 Tank 15 slurry sample (the sample collected during the first bulk waste removal campaign).

Sampling and analysis plan for remediation of Operable Unit 100-IU-3 waste site 600-104

International Nuclear Information System (INIS)

1997-08-01

This sampling and analysis plan (SAP) presents the rationale and strategy for the sampling and analysis activities to support remediation of 100-IU-3 Operable Unit waste site 600-104. The purpose of the proposed sampling and analysis activities is to demonstrate that time-critical remediation of the waste site for soil containing 2,4-Dichlorophonoxyacetic acid salts and esters (2,4-D) and dioxin/furan isomers at concentrations that exceed cleanup levels has been effective. This shall be accomplished by sampling various locations of the waste site before and after remediation, analyzing the samples, and comparing the results to action levels set by the Washington State Department of Ecology

Methods of sampling airborne fungi in working environments of waste treatment facilities.

Science.gov (United States)

Černá, Kristýna; Wittlingerová, Zdeňka; Zimová, Magdaléna; Janovský, Zdeněk

2016-01-01

The objective of the present study was to evaluate and compare the efficiency of a filter based sampling method and a high volume sampling method for sampling airborne culturable fungi present in waste sorting facilities. Membrane filters method was compared with surface air system method. The selected sampling methods were modified and tested in 2 plastic waste sorting facilities. The total number of colony-forming units (CFU)/m3 of airborne fungi was dependent on the type of sampling device, on the time of sampling, which was carried out every hour from the beginning of the work shift, and on the type of cultivation medium (p airborne fungi ranged 2×102-1.7×106 CFU/m3 when using the membrane filters (MF) method, and 3×102-6.4×104 CFU/m3 when using the surface air system (SAS) method. Both methods showed comparable sensitivity to the fluctuations of the concentrations of airborne fungi during the work shifts. The SAS method is adequate for a fast indicative determination of concentration of airborne fungi. The MF method is suitable for thorough assessment of working environment contamination by airborne fungi. Therefore we recommend the MF method for the implementation of a uniform standard methodology of airborne fungi sampling in working environments of waste treatment facilities. This work is available in Open Access model and licensed under a CC BY-NC 3.0 PL license.

Implementation of the waste management transfer act. Requirements from a regulatory point of view; Zur Umsetzung des Entsorgungsuebergangsgesetzes. Anforderungen aus regulatorischer Sicht

Energy Technology Data Exchange (ETDEWEB)

Mueller-Dehn, Christian [PreussenElektra GmbH, Hannover (Germany). Regulierung, Grundsatzfragen

2017-11-15

In future in Germany, the state will be responsible for financing and handling the interim and final storage of radioactive waste from nuclear power plants. With regard to interim storage, this objective is achieved with the provisions of the Waste Management Transfer Act. Regulatory implementation is based on these regulations. BGZ Gesellschaft fuer Zwischenlager mbH is responsible for interim storage on behalf of the Federal Government. Simultaneously with the transfer of interim storage facilities to BGZ a legal transfer of approval is carried out. Insofar as there is a technical, organisational or personnel conjunction with the nuclear power plant operation, which continues to exist beyond this deadline and is relevant for regulatory purposes, a regulation is made via a service contract with the BGZ. This ensures compliance with the licensing regulations. Irradiated fuel assemblies and the waste from reprocessing can be handed over to BGZ from 1 January 2019 onwards and waste with negligible heat generation can be disposed of as of the determination of their proper packaging.

WRAP low level waste (LLW) glovebox operational test report

Energy Technology Data Exchange (ETDEWEB)

Kersten, J.K.

1998-02-19

The Low Level Waste (LLW) Process Gloveboxes are designed to: receive a 55 gallon drum in an 85 gallon overpack in the Entry glovebox (GBIOI); and open and sort the waste from the 55 gallon drum, place the waste back into drum and relid in the Sorting glovebox (GB 102). In addition, waste which requires further examination is transferred to the LLW RWM Glovebox via the Drath and Schraeder Bagiess Transfer Port (DO-07-201) or sent to the Sample Transfer Port (STC); crush the drum in the Supercompactor glovebox (GB 104); place the resulting puck (along with other pucks) into another 85 gallon overpack in the Exit glovebox (GB 105). The status of the waste items is tracked by the Data Management System (DMS) via the Plant Control System (PCS) barcode interface. As an item is moved from the entry glovebox to the exit glovebox, the Operator will track an items location using a barcode reader and enter any required data on the DMS console. The Operational Test Procedure (OTP) will perform evolution`s (described below) using the Plant Operating Procedures (POP) in order to verify that they are sufficient and accurate for controlled glovebox operation.

WRAP low level waste (LLW) glovebox operational test report

International Nuclear Information System (INIS)

Kersten, J.K.

1998-01-01

The Low Level Waste (LLW) Process Gloveboxes are designed to: receive a 55 gallon drum in an 85 gallon overpack in the Entry glovebox (GBIOI); and open and sort the waste from the 55 gallon drum, place the waste back into drum and relid in the Sorting glovebox (GB 102). In addition, waste which requires further examination is transferred to the LLW RWM Glovebox via the Drath and Schraeder Bagiess Transfer Port (DO-07-201) or sent to the Sample Transfer Port (STC); crush the drum in the Supercompactor glovebox (GB 104); place the resulting puck (along with other pucks) into another 85 gallon overpack in the Exit glovebox (GB 105). The status of the waste items is tracked by the Data Management System (DMS) via the Plant Control System (PCS) barcode interface. As an item is moved from the entry glovebox to the exit glovebox, the Operator will track an items location using a barcode reader and enter any required data on the DMS console. The Operational Test Procedure (OTP) will perform evolution's (described below) using the Plant Operating Procedures (POP) in order to verify that they are sufficient and accurate for controlled glovebox operation

Portable sensor for hazardous waste

International Nuclear Information System (INIS)

Piper, L.G.

1994-01-01

Objective was to develop a field-portable monitor for sensitive hazardous waste detection using active nitrogen energy transfer (ANET) excitation of atomic and molecular fluorescence (active nitrogen is made in a dielectric-barrier discharge in nitrogen). It should provide rapid field screening of hazardous waste sites to map areas of greatest contamination. Results indicate that ANET is very sensitive for monitoring heavy metals (Hg, Se) and hydrocarbons; furthermore, chlorinated hydrocarbons can be distinguished from nonchlorinated ones. Sensitivity is at ppB levels for sampling in air. ANET appears ideal for on-line monitoring of toxic heavy metal levels at building sites, hazardous waste land fills, in combustor flues, and of chlorinated hydrocarbon levels at building sites and hazardous waste dumps

ESTIMATING HIGH LEVEL WASTE MIXING PERFORMANCE IN HANFORD DOUBLE SHELL TANKS

International Nuclear Information System (INIS)

Thien, M.G.; Greer, D.A.; Townson, P.

2011-01-01

The ability to effectively mix, sample, certify, and deliver consistent batches of high level waste (HLW) feed from the Hanford double shell tanks (DSTs) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. The Department of Energy's (DOE's) Tank Operations Contractor (TOC), Washington River Protection Solutions (WRPS) is currently demonstrating mixing, sampling, and batch transfer performance in two different sizes of small-scale DSTs. The results of these demonstrations will be used to estimate full-scale DST mixing performance and provide the key input to a programmatic decision on the need to build a dedicated feed certification facility. This paper discusses the results from initial mixing demonstration activities and presents data evaluation techniques that allow insight into the performance relationships of the two small tanks. The next steps, sampling and batch transfers, of the small scale demonstration activities are introduced. A discussion of the integration of results from the mixing, sampling, and batch transfer tests to allow estimating full-scale DST performance is presented.

Statistical Methods and Tools for Hanford Staged Feed Tank Sampling

Energy Technology Data Exchange (ETDEWEB)

Fountain, Matthew S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Brigantic, Robert T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Peterson, Reid A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2013-10-01

This report summarizes work conducted by Pacific Northwest National Laboratory to technically evaluate the current approach to staged feed sampling of high-level waste (HLW) sludge to meet waste acceptance criteria (WAC) for transfer from tank farms to the Hanford Waste Treatment and Immobilization Plant (WTP). The current sampling and analysis approach is detailed in the document titled Initial Data Quality Objectives for WTP Feed Acceptance Criteria, 24590-WTP-RPT-MGT-11-014, Revision 0 (Arakali et al. 2011). The goal of this current work is to evaluate and provide recommendations to support a defensible, technical and statistical basis for the staged feed sampling approach that meets WAC data quality objectives (DQOs).

BY tank farm waste inventory and transfer data for ITS-2 operation during January To December 1971

Energy Technology Data Exchange (ETDEWEB)

Reich, F.R., Westinghouse Hanford

1996-08-02

Data record inventory of pumping activities and liquid level changes including occasional operations comments for the BY Tank Farm. Waste inventory and transfer data for ITS-2 operation during January to December 1971.

Characterization of Hanford waste and the role of historic modeling

International Nuclear Information System (INIS)

Simpson, B.C.; Eberlein, S.J.; Brown, T.M.; Brevick, C.H.; Angew, S.F.

1996-02-01

The tank waste characterization process is an integral part of the overall effort to identify, quantify and control the hazards associated with radioactive wastes stored in underground tanks at the Hanford Reservation. Characterization of the current waste tank contents through the use of waste sampling is only partly effective. The historic records must be exploited as much as possible. A model generates an estimate of the current contents of each tank, built up from the estimated volumes of each of the defined waste components. The model combines the best estimate of the waste stream composition for each of the major waste generating processes. All available waste transfer records were compiled and integrated to track waste tank fill history. The behavior of the waste materials in the tanks was modeled, based on general scientific principles augmented with specific measurement data. Sample analysis results were not used directly to generate any of the tank contents estimates, but were used to determine the values of variable parameters such as the solubility. By considering all available information first (including historical model estimates, surveillance data, and past sample analysis results), future sampling resources and other characterization efforts can best be spent on tanks that will provide the largest returns of information

Determination of Intrinsic Permeability for Packed Waste of Indonesian Solid Waste

Directory of Open Access Journals (Sweden)

Benno Rahardyan

2010-11-01

Full Text Available Gas permeability and intrinsic permeability are the major parameters to promote aeration for packed waste. The objectives of this research are to identify physical parameters of gas transfer from a various type of packed wastes and examine ventilation design theory for landfill to enhance waste stabilization. Method to determine value of gas permeability and intrinsic permeability for packed waste is by flushing the packed column containing various type and physical characteristics of wastes with an air pump. Permeability was calculated by measuring pressure gradient on sampling points of the column using inclined manometer at distance 10 cm, 23 cm, 46 cm, 69 cm, 92 cm and 115 cm from origin. Gas permeability is specifically relied on physical parameters of wastes as follows, density, moisture content, particle size and gas velocity on the surface of compacted waste layer. Compost has finer pore structure and smaller pore size than leaves as well as mixed organic (65% and inorganic wastes (35%. The experiment found the intrinsic permeability of leaves waste are in the order of 10-11 to 10-8 m2, 10-11 to 10-9 m2 for compost and 10-9 m2 for mixed organic (65% and inorganic wastes (35%.

Quality assurance procedures for the analysis of TRU waste samples

International Nuclear Information System (INIS)

Glasgow, D.C. Giaquinto, J.M.; Robinson, L.

1995-01-01

The Waste Isolation Pilot Plant (WIPP) project was undertaken in response to the growing need for a national repository for transuranic (TRU) waste. Guidelines for WIPP specify that any waste item to be interred must be fully characterized and analyzed to determine the presence of chemical compounds designated hazardous and certain toxic elements. The Transuranic Waste Characterization Program (TWCP) was launched to develop analysis and quality guidelines, certify laboratories, and to oversee the actual waste characterizations at the laboratories. ORNL is participating in the waste characterization phase and brings to bear a variety of analytical techniques including ICP-AES, cold vapor atomic absorption, and instrumental neutron activation analysis (INAA) to collective determine arsenic, cadmium, barium, chromium, mercury, selenium, silver, and other elements. All of the analytical techniques involved participate in a cooperative effort to meet the project objectives. One important component of any good quality assurance program is determining when an alternate method is more suitable for a given analytical problem. By bringing to bear a whole arsenal of analytical techniques working toward common objectives, few analytical problems prove to be insurmountable. INAA and ICP-AES form a powerful pair when functioning in this cooperative manner. This paper will provide details of the quality assurance protocols, typical results from quality control samples for both INAA and ICP-AES, and detail method cooperation schemes used

A study on the representative sampling survey for the inspection of the clearance level for the radioisotope waste

International Nuclear Information System (INIS)

Hong Joo Ahn; Se Chul Sohn; Kwang Yong Jee; Ju Youl Kim; In Koo Lee

2007-01-01

Utilization facilities for radioisotope (RI) are increasing annually in South Korea, and the total number was 2,723, as of December 31, 2005. The inspection of a clearance level is a very important problem in order to ensure a social reliance for releasing radioactive materials to the environment. Korean regulations for such a clearance are described in Notice No. 2001-30 of the Ministry of Science and Technology (MOST) and Notice No. 2002-67 of the Ministry of Commerce, Industry and Energy (MOCIE). Most unsealed sources in RI waste drums at a storage facility are low-level beta-emitters with short half-lives, so it is impossible to measure their inventories by a nondestructive analysis. Furthermore, RI wastes generated from hospital, educational and research institutes and industry have a heterogeneous, multiple, irregular, and a small quantity of a waste stream. This study addresses a representative (master) sampling survey and analysis plan for RI wastes because a complete enumeration of waste drums is impossible and not desirable in terms of a cost and efficiency. The existing approaches to a representative sampling include a judgmental, simple random, stratified random, systematic grid, systematic random, composite, and adaptive sampling. A representative sampling plan may combine two or more of the above sampling approaches depending on the type and distribution of a waste stream. Stratified random sampling (constrained randomization) is proven to be adequate for a sampling design of a RI waste regarding a half-life, surface dose, undertaking time to a storage facility, and type of waste. The developed sampling protocol includes estimating the number of drums within a waste stream, estimating the number of samples, and a confirmation of the required number of samples. The statistical process control for a quality assurance plan includes control charts and an upper control limit (UCL) of 95% to determine whether a clearance level is met or not. (authors)

Modulation transfer function cascade model for a sampled IR imaging system.

Science.gov (United States)

de Luca, L; Cardone, G

1991-05-01

The performance of the infrared scanning radiometer (IRSR) is strongly stressed in convective heat transfer applications where high spatial frequencies in the signal that describes the thermal image are present. The need to characterize more deeply the system spatial resolution has led to the formulation of a cascade model for the evaluation of the actual modulation transfer function of a sampled IR imaging system. The model can yield both the aliasing band and the averaged modulation response for a general sampling subsystem. For a line scan imaging system, which is the case of a typical IRSR, a rule of thumb that states whether the combined sampling-imaging system is either imaging-dependent or sampling-dependent is proposed. The model is tested by comparing it with other noncascade models as well as by ad hoc measurements performed on a commercial digitized IRSR.

A tracking system for groundwater sampling and data transfer schedules

International Nuclear Information System (INIS)

Mercier, T.M.

1990-12-01

Since groundwater monitoring programs at the Oak Ridge Y-12 Plant have become more complex and varied and as the occasions to respond to internal and external reporting requirements have become more frequent and time constrained, the need to track groundwater sampling activities and data transfer from the analytical laboratories has become imperative. If backlogs can be caught early, resources can be added or reallocated in the field and in the laboratory in a timely manner to ensure reporting deadlines are met. The tracking system discussed in this paper starts with clear definition of the groundwater monitoring program at the facility. This information is input into base datasets at the beginning of the sampling cycle. As the sampling program progresses, information about well sampling dates and data transfer dates is input into the base datasets. From the base program data and the update data, a status report is periodically generated by a computer program which identifies the type and nature of bottle necks encountered during the implementation of the groundwater monitoring program

Emerging halogenated flame retardants and hexabromocyclododecanes in food samples from an e-waste processing area in Vietnam.

Science.gov (United States)

Tao, Fang; Matsukami, Hidenori; Suzuki, Go; Tue, Nguyen Minh; Viet, Pham Hung; Takigami, Hidetaka; Harrad, Stuart

2016-03-01

This study reports concentrations of selected emerging halogenated flame retardants (HFRs) and hexabromocyclododecanes (HBCDs) in foodstuffs sourced from an e-waste processing area in Vietnam and two reference sites in Vietnam and Japan. Concentrations of all target HFRs in e-waste-impacted samples in this study exceed significantly (p e-waste processing activities exert a substantial impact on local environmental contamination and human dietary exposure. Significant linear positive correlations in concentrations of syn-Dechlorane Plus (DP) and anti-DP were found between soils and those in co-located chicken samples (p e-waste processing sites and non-e-waste processing areas elsewhere.

The health services wastes management of a sample of brazilian hospitals

Directory of Open Access Journals (Sweden)

Claude Machline

2006-07-01

Full Text Available This paper focuses the Health Services wastes management of 70 Brazilian hospitals. As the outcome of a distance course, in 2003, each hospital was required to describe its existing Health Services wastes system and its Plan for improvement.The project was administered by an association of two leading Brazilian educational entities, the Fundação Getulio Vargas and the Universidade Federal de Santa Catarina. Data concerning collection, disposal and final treatment of infectious, hazardous, chemical, radioactive and common wastes were tabulated and analysed. Water supply, liquid effluents and gaseous emissions were also investigated..Their technical and economical aspects were appraised. The research indicates that the sampled hospitals are still in an incipient stage of wastes management. An extensive gap exists between the present situation and the legal and acceptable requirements they should comply with, both on health care and on environmental standpoints.

Governmental Decree No. 224/1997 of 13 August 1997 on fees levied on nuclear waste producers and on the transfer of such fees to the Nuclear Account

International Nuclear Information System (INIS)

1997-01-01

Nuclear power plants are obliged to pay a fee of CZK 50 per MWe produced. This amount is transferred to the Nuclear Account. The fee return is filed with the Czech Radioactive Repositories Administration. Minor radioactive waste producers only pay the radioactive waste storage cost; this sum is also transferred to the Nuclear Account. Each entity applying with the State Office for Nuclear Safety for a licence regarding activities which are associated with the production of radioactive wastes shall register with the Radioactive Repositories Administration as a prospective radioactive waste producer. (P.A.)

Characterization and Leach Testing for REDOX Sludge and S-Saltcake Actual Waste Sample Composites

International Nuclear Information System (INIS)

Fiskum, Sandra K.; Buck, Edgar C.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Hubler, Timothy L.; Jagoda, Lynette K.; Jenson, Evan D.; Kozelisky, Anne E.; Lumetta, Gregg J.; MacFarlan, Paul J.; McNamara, Bruce K.; Peterson, Reid A.; Sinkov, Sergey I.; Snow, Lanee A.; Swoboda, Robert G.

2008-01-01

This report describes processing and analysis results of boehmite waste type (Group 5) and insoluble high Cr waste type (Group 6). The sample selection, compositing, subdivision, physical and chemical characterization are described. Extensive batch leach testing was conducted to define kinetics and leach factors of selected analytes as functions of NaOH concentration and temperature. Testing supports issue M-12 resolution for the Waste Treatment Plant

Characterization and Leach Testing for REDOX Sludge and S-Saltcake Actual Waste Sample Composites

Energy Technology Data Exchange (ETDEWEB)

Fiskum, Sandra K.; Buck, Edgar C.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Hubler, Timothy L.; Jagoda, Lynette K.; Jenson, Evan D.; Kozelisky, Anne E.; Lumetta, Gregg J.; MacFarlan, Paul J.; McNamara, Bruce K.; Peterson, Reid A.; Sinkov, Sergey I.; Snow, Lanee A.; Swoboda, Robert G.

2008-07-10

This report describes processing and analysis results of boehmite waste type (Group 5) and insoluble high Cr waste type (Group 6). The sample selection, compositing, subdivision, physical and chemical characterization are described. Extensive batch leach testing was conducted to define kinetics and leach factors of selected analytes as functions of NaOH concentration and temperature. Testing supports issue M-12 resolution for the Waste Treatment Plant.

Method for fractional solid-waste sampling and chemical analysis

DEFF Research Database (Denmark)

Riber, Christian; Rodushkin, I.; Spliid, Henrik

2007-01-01

four subsampling methods and five digestion methods, paying attention to the heterogeneity and the material characteristics of the waste fractions, it was possible to determine 61 substances with low detection limits, reasonable variance, and high accuracy. For most of the substances of environmental...... of variance (20-85% of the overall variation). Only by increasing the sample size significantly can this variance be reduced. The accuracy and short-term reproducibility of the chemical characterization were good, as determined by the analysis of several relevant certified reference materials. Typically, six...... to eight different certified reference materials representing a range of concentrations levels and matrix characteristics were included. Based on the documentation provided, the methods introduced were considered satisfactory for characterization of the chemical composition of waste-material fractions...

Defense-in-depth evaluation for the New Waste Transfer Facility

International Nuclear Information System (INIS)

Hayes, T.G.; Kelly, J.L.

1995-01-01

This report fulfills part of the requirements of References 2 and 3 by documenting a Defense-In-Depth evaluation for the New Waste Transfer Facility (NWTF). This evaluation was performed using methodology similar to that used in an evaluation for the Defense Waste Processing Facility (DWPF). It differs because the DWPF evaluation was based on an existing Process Hazards Analysis (PHA) while NWTF's is based on a Preoperational Process Hazards Review (PHR) (Ref. 1). The accidents in the Process Hazards Review (PHR) were reviewed to determine those that might have significant consequences. Significance was based on the findings of the PHR, The facility design was reviewed to determine the Structures, Systems, and Components (SSCs) and administrative controls available before and after each accident. From this was developed a list of the Lines of Defense (LODs) available to contain the hazard associated with the accident. A summary of these LODs is given in Appendix C. Items are tabulated that are suggested for consideration in the functional classification as worker protection items. The specific criteria used in the evaluation is given in the methodology section of this report. The results are documented in Appendices A, B, C, and D

Hazard evaluation for transfer of waste from tank 241-SY-101 to tank 241-SY-102

International Nuclear Information System (INIS)

SHULTZ, M.V.

1999-01-01

Tank 241-SY-101 (SY-101) waste level growth is an emergent, high priority issue. The purpose of this document is to record the hazards evaluation process and document potential hazardous conditions that could lead to the release of radiological and toxicological material from the proposed transfer of a limited quantity (approximately 100,000 gallons) of waste from SY-101 to 241-SY-102 (SY-102). The results of the hazards evaluation will be compared to the current Tank Waste Remediation System (TWRS) Basis for Interim Operation (HNF-SD-WM-BIO-001, 1998, Revision 1) to identify any hazardous conditions where Authorization Basis (AB) controls may not be sufficient or may not exist. Comparison to LA-UR-92-3196, A Safety Assessment for Proposed Pump Mixing Operations to Mitigate Episodic Gas Releases in Tank 241-SY-101, was also made in the case of transfer pump removal activities. This document is not intended to authorize the activity or determine the adequacy of controls; it is only intended to provide information about the hazardous conditions associated with this activity. The Unreviewed Safety Question (USQ) process will be used to determine the adequacy of controls and whether the proposed activity is within the AB. This hazard evaluation does not constitute an accident analysis

EG ampersand G Mount Plant, December 1990 and January 1991, D ampersand D soil box sampling

International Nuclear Information System (INIS)

1991-04-01

Six hundred eighty-two (682) containers of soil were generated at Mound Plant between April 1 and October 31, 1990 as a result of the excavation of soils containing plutonium-238 at two ongoing Decontamination and Decommissioning (D ampersand D) Program sites; these areas are known as Area 14, the waste transfer system (WTS) hillside, and Area 17, the Special Metallurgical (SM) Building Area. The soils from these areas are part of the Mound Plant waste stream number AMDM-000000010, Contaminated Soil, and are proposed for shipment to the Nevada Test Site (NTS) for disposal as low-level radioactive waste. These containers of soil are currently in storage at Mound Plant. The purpose of this sampling and analysis was to demonstrate that the D ampersand D soils comply with the waste acceptance requirements of the NTS, as presented In Nevada Test Site Defense Waste Acceptance Criteria, Certification, and Transfer Requirements (DOE 1988). The sealed waste packages, constructed of wood or metal, are currently being stored In Building 31 and at other locations throughout the Mound Plant. For additional historical information concerning the D ampersand D soils, Including waste stream evaluations and past sampling data see the Sampling and Analysis Plan for Mound Plant D ampersand D Soils Packages (EG ampersand G 1991)

Determination of Different Metals in Steel Waste Samples Using laser Induced Breakdown Spectroscopy

Directory of Open Access Journals (Sweden)

A. H. Bakry

2007-12-01

Full Text Available Elemental analysis of waste samples collected from steel products manufacturing plant (SPS located at industrial city of Jeddah, Saudi-Arabia has been carried out using Laser Induced Breakdown Spectroscopy (LIBS. The 1064 nm laser radiations from a Nd:YAG laser at an irradiance of 7.6  1010 W cm –2 were used. Atomic emission spectra of the elements present in the waste samples were recorded in the 200 – 620 nm region. Elements such as Fe, W, Ti, Al, Mg, Ca, S, Mn, and Na were detected in these samples. Quantitative determination of the elemental concentration was obtained for these metals against certified standard samples. Parametric dependences of LIBS signal intensity on incident laser energy and time delay between the laser pulse and data acquisition system were also carried out.

Review of Analytes of Concern and Sample Methods for Closure of DOE High Level Waste Storage Tanks

International Nuclear Information System (INIS)

Thomas, T.R.

2002-01-01

Sampling residual waste after tank cleaning and analysis for analytes of concern to support closure and cleaning targets of large underground tanks used for storage of legacy high level radioactive waste (HLW) at Department of Energy (DOE) sites has been underway since about 1995. The DOE Tanks Focus Area (TFA) has been working with DOE tank sites to develop new sampling plans, and sampling methods for assessment of residual waste inventories. This paper discusses regulatory analytes of concern, sampling plans, and sampling methods that support closure and cleaning target activities for large storage tanks at the Hanford Site, the Savannah River Site (SRS), the Idaho National Engineering and Environmental Laboratory (INEEL), and the West Valley Demonstration Project (WVDP)

Genotoxicity studies in semiconductor industry. 1. In vitro mutagenicity and genotoxicity studies of waste samples resulting from plasma etching

Energy Technology Data Exchange (ETDEWEB)

Braun, R.; Huettner, E.M.; Merten, H.; Raabe, F. (Institute of Plant Genetics and Crop Plant Research, Gatersleben (Germany))

1993-07-01

Solid waste samples taken from the etching reactor, the turbo pump, and the waste air system of a plasma etching technology line in semiconductor production were studied as to their genotoxic properties in a bacterial repair test, in the Ames/Salmonella microsome assay, in the SOS chromotest, in primary mouse hepatocytes, and in Chinese hamster V79 cell cultures. All three waste samples were found to be active by inducing of unscheduled DNA-synthesis in mouse hepatocytes in vitro. In the bacterial rec-type repair test with Proteus mirabilis, waste samples taken from the turbo pump and the vacuum pipe system were not genotoxic. The waste sample taken from the chlorine-mediated plasma reactor was clearly positive in the bacterial repair assay and in the SOS chromotest with Escherichia coli. Mutagenic activity was demonstrated for all samples in the presence and absence of S9 mix made from mouse liver homogenate. Again, highest mutagenic activity was recorded for the waste sample taken from the plasma reactor, while samples collected from the turbo pump and from the waste air system before dilution and liberation of the air were less mutagenic. For all samples chromosomal damage in V79 cells was not detected, indicating absence of clastogenic activity in vitro. Altogether, these results indicate generation of genotoxic and mutagenic products as a consequence of chlorine-mediated plasma etching in the microelectronics industry and the presence of genotoxins even in places distant from the plasma reactor. Occupational exposure can be expected both from the precipitated wastes and from chemicals reaching the environment with the air stream.

Data Quality Objectives for Regulatory Requirements for Dangerous Waste Sampling and Analysis

International Nuclear Information System (INIS)

MULKEY, C.H.

1999-01-01

This document describes sampling and analytical requirements needed to meet state and federal regulations for dangerous waste (DW). The River Protection Project (RPP) is assigned to the task of storage and interim treatment of hazardous waste. Any final treatment or disposal operations, as well as requirements under the land disposal restrictions (LDRs), fall in the jurisdiction of another Hanford organization and are not part of this scope. The requirements for this Data Quality Objective (DQO) Process were developed using the RPP Data Quality Objective Procedure (Banning 1996), which is based on the U.S. Environmental Protection Agency's (EPA) Guidance for the Data Quality Objectives Process (EPA 1994). Hereafter, this document is referred to as the DW DQO. Federal and state laws and regulations pertaining to waste contain requirements that are dependent upon the composition of the waste stream. These regulatory drivers require that pertinent information be obtained. For many requirements, documented process knowledge of a waste composition can be used instead of analytical data to characterize or designate a waste. When process knowledge alone is used to characterize a waste, it is a best management practice to validate the information with analytical measurements

Assessing sample attenuation parameters for use in low-energy efficiency transfer in gamma-ray spectrometry

International Nuclear Information System (INIS)

Bruggeman, M.; Verheyen, L.; Vidmar, T.; Liu, B.

2016-01-01

We present a numerical fitting method for transmission data that outputs an equivalent sample composition. This output is used as input to a generalised efficiency transfer model based on the EFFTRAN software integrated in a LIMS. The procedural concept allows choosing between efficiency transfer with a predefined sample composition or with an experimentally determined composition based on a transmission measurement. The method can be used for simultaneous quantification of low-energy gamma emitters like "2"1"0Pb, "2"4"1Am, "2"3"4Th in typical environmental samples. - Highlights: • New fitting method for experimentally determined attenuation coefficients. • Generalised efficiency transfer with EFFTRAN based on transmission measurements. • Method of generalized efficiency transfer integrated in LIMS. • Method applicable to gamma-ray spectrometry of environmental samples.

Off-Gas Analysis During the Vitrification of Hanford Radioactive Waste Samples

International Nuclear Information System (INIS)

Ha, B.C.; Ferrara, D.M.; Crawford, C.L.; Choi, A.S.; Bibler, N.E.

1998-01-01

This paper describes the off-gas analysis of samples collected during the radioactive vitrification experiments. Production and characterization of the Hanford waste-containing LAW and HAW glasses are presented in related reports from this conference

Heat transfer in vitrified radioactive waste

International Nuclear Information System (INIS)

Palancar, M.C.; Luis, M.A.; Luis, P.; Aragon, J.M.; Montero, M.A.

1987-01-01

An experimental method for measuring the thermal conductivity and convection coefficient of borosilicate glass cylinders, containing a simulated high level radioactive waste, is described. A simulation of the thermal behaviour of matrices of solidified waste during the cooling in air, water and a geological repository has been done. The experimental values of the thermal conductivity are ranging from 0.267 to 0.591 w/m K, for matrices with simulated waste contents of 10 to 40% (the waste is simulated by no radioactive isotopes). The convection coefficient for air/cylinders under the operating conditions used is 116 w/m 2 K. The simulated operation of cooling in air shows that about 1-2 days are enough to cool a solidified waste cylinder 0.6m diameter from 900 to 400 0 C. The cooling under water from 400 to near 80 0 C is faster than in air, but sharp temperature gradients within the matrices could be expected. The simulation of geological repositories lead to some criteria of arranging the matrices for avoiding undesirable high temperature points. (author) 1 fig

Fabrication of a Sludge-Conditioning System for processing legacy wastes from the Gunite and Associated Tanks

International Nuclear Information System (INIS)

Randolph, J.D.; Lewis, B.E.; Farmer, J.R.; Johnson, M.A.

2000-01-01

The Sludge Conditioning System (SCS) for the Gunite and Associated Tanks (GAATs) is designed to receive, monitor, characterize and process legacy waste materials from the South Tank Farm tanks in preparation for final transfer of the wastes to the Melton Valley Storage Tanks (MVSTs), which are located at Oak Ridge National Laboratory. The SCS includes (1) a Primary Conditioning System (PCS) Enclosure for sampling and particle size classification, (2) a Solids Monitoring Test Loop (SMTL) for slurry characterization, (3) a Waste Transfer Pump to retrieve and transfer waste materials from GAAT consolidation tank W-9 to the MVSTs, (4) a PulsAir Mixing System to provide mixing of consolidated sludges for ease of retrieval, and (5) the interconnecting piping and valving. This report presents the design, fabrication, cost, and fabrication schedule information for the SCS

Methodology of testing environmental samples from the area surrounding radioactive waste deposits

International Nuclear Information System (INIS)

Kropikova, S.; Pastuchova, D.

1979-01-01

Methods are described of environmental sample investigation in the area surrounding radioactive waste deposits, namely monitoring ground water, surface water, sediments, water flows and catchments, vegetation and soil. Methods of sample preparation, and methods of radionuclides determination in mixtures are also discussed, as are spot activity measurement methods. (author)

Thermal modeling of core sampling in flammable gas waste tanks. Part 2: Rotary-mode sampling

International Nuclear Information System (INIS)

Unal, C.; Poston, D.; Pasamehmetoglu, K.O.; Witwer, K.S.

1997-01-01

The radioactive waste stored in underground storage tanks at Hanford site includes mixtures of sodium nitrate and sodium nitrite with organic compounds. The waste can produce undesired violent exothermic reactions when heated locally during the rotary-mode sampling. Experiments are performed varying the downward force at a maximum rotational speed of 55 rpm and minimum nitrogen purge flow of 30 scfm. The rotary drill bit teeth-face temperatures are measured. The waste is simulated with a low thermal conductivity hard material, pumice blocks. A torque meter is used to determine the energy provided to the drill string. The exhaust air-chip temperature as well as drill string and drill bit temperatures and other key operating parameters were recorded. A two-dimensional thermal model is developed. The safe operating conditions were determined for normal operating conditions. A downward force of 750 at 55 rpm and 30 scfm nitrogen purge flow was found to yield acceptable substrate temperatures. The model predicted experimental results reasonably well. Therefore, it could be used to simulate abnormal conditions to develop procedures for safe operations

Ratio methods for cost-effective field sampling of commercial radioactive low-level wastes

International Nuclear Information System (INIS)

Eberhardt, L.L.; Simmons, M.A.; Thomas, J.M.

1985-07-01

In many field studies to determine the quantities of radioactivity at commercial low-level radioactive waste sites, preliminary appraisals are made with field radiation detectors, or other relatively inaccurate devices. More accurate determinations are subsequently made with procedures requiring chemical separations or other expensive analyses. Costs of these laboratory determinations are often large, so that adequate sampling may not be achieved due to budget limitations. In this report, we propose double sampling as a way to combine the expensive and inexpensive aproaches to substantially reduce overall costs. The underlying theory was developed for human and agricultural surveys, and is partially based on assumptions that are not appropriate for commercial low-level waste sites. Consequently, extensive computer simulations were conducted to determine whether the results can be applied in circumstances of importance to the Nuclear Regulatory Commission. This report gives the simulation details, and concludes that the principal equations are appropriate for most studies at commercial low-level waste sites. A few points require further research, using actual commercial low-level radioactive waste site data. The final section of the report provides some guidance (via an example) for the field use of double sampling. Details of the simulation programs are available from the authors. Major findings are listed in the Executive Summary. 9 refs., 9 figs., 30 tabs

W-320 waste retrieval sluicing system transfer line flushing volume and frequency calculation

International Nuclear Information System (INIS)

Bailey, J.W.

1997-01-01

The calculations contained in this analysis document establish the technical basis for the volume, frequency, and flushing fluid to be utilized for routine Waste Retrieval Sluicing System (WRSS) process line flushes. The WRSS was installed by Project W-320, Tank 241-C-106 Sluicing. The double contained pipelines being flushed have 4 inch stainless steel primary pipes. The flushes are intended to prevent hydrogen buildup in the transfer lines and to provide ALARA conditions for maintenance personnel

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.

Tank vapor sampling and analysis data package for tank 241-C-106 waste retrieval sluicing system process test phase III, sampled March 28, 1999

International Nuclear Information System (INIS)

LOCKREM, L.L.

1999-01-01

This data package presents sampling data and analytical results from the March 28, 1999, vapor sampling of Hanford Site single-shell tank 241-C-106 during active sluicing. Samples were obtained from the 296-C-006 ventilation system stack and ambient air at several locations. Characterization Project Operations (CPO) was responsible for the collection of all SUMMATM canister samples. The Special Analytical Support (SAS) vapor team was responsible for the collection of all triple sorbent trap (TST), sorbent tube train (STT), polyurethane foam (PUF), and particulate filter samples collected at the 296-C-006 stack. The SAS vapor team used the non-electrical vapor sampling (NEVS) system to collect samples of the air, gases, and vapors from the 296-C-006 stack. The SAS vapor team collected and analyzed these samples for Lockheed Martin Hanford Corporation (LMHC) and Tank Waste Remediation System (TWRS) in accordance with the sampling and analytical requirements specified in the Waste Retrieval Sluicing System Vapor Sampling and Analysis Plan (SAP) for Evaluation of Organic Emissions, Process Test Phase III, HNF-4212, Rev. 0-A, (LMHC, 1999). All samples were stored in a secured Radioactive Materials Area (RMA) until the samples were radiologically released and received by SAS for analysis. The Waste Sampling and Characterization Facility (WSCF) performed the radiological analyses. The samples were received on April 5, 1999

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

Characterization of the first core sample of neutralized current acid waste from double-shell tank 101-AZ

International Nuclear Information System (INIS)

Peterson, M.E.; Scheele, R.D.; Tingey, J.M.

1989-09-01

In FY 1989, Westinghouse Hanford Company (WHC) successfully obtained four core samples (totaling seven segments) of neutralized current acid waste (NCAW) from double-shell tanks (DSTs) 101-AZ and 102-AZ. A segment was a 19-in.-long and 1-in.-diameter cylindrical sample of waste. A core sample consisted of enough 19-in.-long segments to obtain the waste of interest. Three core samples were obtained from DST 101-AZ and one core sample from DST 102-AZ. Two DST 101-AZ core samples consisted of two segments per core, and the third core sample consisted of only one segment. The third core consisted of the solids from the bottom of the tank and was used to determine the relative abrasiveness of this NCAW. The DST 102-AZ core sample consisted of two segments. The core samples were transported to the Pacific Northwest Laboratory (PNL), where the waste was extruded from its sampler and extensively characterized. A characterization plan was followed that simulated the processing of the NCAW samples through retrieval, pretreatment and vitrification process steps. Physical, rheological, chemical and radiochemical properties were measured throughout the process steps. The characterization of the first core sample from DST 101-AZ was completed, and the results are provided in this report. The results for the other core characterizations will be reported in future reports. 3 refs., 13 figs., 10 tabs

Sandia National Laboratories Mixed Waste Landfill Integrated Demonstration

International Nuclear Information System (INIS)

Tyler, L.D.; Phelan, J.M.; Prindle, N.K.; Purvis, S.T.; Stormont, J.C.

1992-01-01

The Mixed-Waste Landfill Integrated Demonstration (MWLID) has been assigned to Sandia National Laboratories (SNL) by the US Department of Energy (DOE) Office of Technology Development. The mission of the MWLID is to assess, implement and transfer technologies and systems that lead to quicker, safer, and more efficient remediation of buried chemical and mixed-waste sites. The MWLID focus is on two landfills at SNL in Albuquerque, New Mexico: The Chemical Waste Landfill (CWL) and the Mixed-Waste Landfill (MWL). These landfills received chemical, radioactive and mixed wastes from various SNL nuclear research programs. A characterization system has been designed for the definition of the extent and concentration of contamination. This system includes historical records, directional drilling, and emplacement membrane, sensors, geophysics, sampling strategy, and on site sample analysis. In the remediation task, in-situ remediation systems are being designed to remove volatile organic compounds (VOC's) and heavy metals from soils. The VOC remediation includes vacuum extraction with electrical and radio-frequency heating. For heavy metal contamination, electrokinetic processes are being considered. The MWLID utilizes a phased, parallel approach. Initial testing is performed at an uncontaminated site adjacent to the CWL. Once characterization is underway at the CWL, lessons learned can be directly transferred to the more challenging problem of radioactive waste in the MWL. The MWL characterization can proceed in parallel with the remediation work at CWL. The technologies and systems demonstrated in the MWLID are to be evaluated based on their performance and cost in the real remediation environment of the landfills

[PHAHs levels in soil samples from the E-waste disassembly sites and their sources allocation].

Science.gov (United States)

Zhao, Gao-Feng; Wang, Zi-Jian

2009-06-15

Soil samples (each with 3 replicates of - 1 kg, at the top 0-5 cm layer) were collected from each of the e-waste disassembly sites and the control site. Also obtained from each disassembly site were samples (each weighing - 0.2 kg) of cable coating,stuffing powder, and circuit boards chipping. The contents of 23 PBB congeners, 12 PBDE congeners, and 27 PCB congeners in soil and in their potential sources, including e-waste residues, were measured using the GC-MS5975B technique. The highest level of PBBs was found in the cable coating among the three e-waste residues, with a concentration of 35.25 ng x g(-1). The contents of low-brominated PBBs (including monobromobiphenyls and dibromobiphenyls) accounted for 38% of the total PBBs concentration observed in cable coating sample. The highest levels of PBDEs and PBDE209 were found in the stuffing powder for electronic component among the collected e-waste residues, with a concentration of 29.71 and 4.19 x 10(3) ng x g(-1). PBDE153 and PBDE183 were the most predominant PBDE congeners, with their concentration accounting for 43% and 24% of the total PBDEs concentration observed in the stuffing powder sample, respectively. Levels of PCBs in cable coating were the highest in these e-waste residues, with a concentration of 680.02 ngx g(-1). The observed values of the three PHAHs in soils from the disassembly site were considerably higher than their corresponding values observed in the control site (p < 0.05), which indicates that these PHAHs from e-waste is the pollution source of local environment.

Sampling and analysis plan for remediation of Operable Unit 100-IU-3 waste site 600-104. Revision 1

International Nuclear Information System (INIS)

1997-08-01

This sampling and analysis plan presents the rationale and strategy for the sampling and analysis activities to support remediation of 100-IU-3 Operable Unit waste site 600-104. The purpose of the proposed sampling and analysis activities is to demonstrate that time-critical remediation of the waste site for soil containing 2,4-Dichlorophenoxyacetic acid salts and esters (2,4-D) and dioxin/furan isomers at concentrations that exceed cleanup levels has been effective. This shall be accomplished by sampling various locations of the waste site before and after remediation, analyzing the samples, and comparing the results to action levels set by the Washington State Department of Ecology

Listed waste history at Hanford facility TSD units

International Nuclear Information System (INIS)

Miskho, A.G.

1996-01-01

This document was prepared to close out an occurrence report that Westinghouse Hanford Company issued on December 29, 1994. Occurrence Report RL-WHC-GENERAL-1994-0020 was issued because knowledge became available that could have impacted start up of a Hanford Site facility. The knowledge pertained to how certain wastes on the Hanford Site were treated, stored, or disposed of. This document consolidates the research performed by Westinghouse Hanford Company regarding listed waste management at onsite laboratories that transfer waste to the Double-Shell Tank System. Liquid and solid (non-liquid) dangerous wastes and mixed wastes at the Hanford Site are generated from various Site operations. These wastes may be sampled and characterized at onsite laboratories to meet waste management requirements. In some cases, the wastes that are generated in the field or in the laboratory from the analysis of samples require further management on the Hanford Site and are aggregated together in centralized tank storage facilities. The process knowledge presented herein documents the basis for designation and management of 242-A Evaporator Process Condensate, a waste stream derived from the treatment of the centralized tank storage facility waste (the Double-Shell Tank System). This document will not be updated as clean up of the Hanford Site progresses

Radionuclide mass transfer rates from a pinhole in a waste container for an inventory-limited and a constant concentration source

International Nuclear Information System (INIS)

LeNeveu, D.M.

1996-03-01

Analytical solutions for transient and steady state diffusive mass transfer rates from a pinhole in a waste container are developed for constant concentration and inventory-limited source conditions. Mass transport in three media are considered, inside the pinhole (medium 2), outside the container (medium 3) and inside the container (medium 1). Simple equations are developed for radionuclide mass transfer rates from a pinhole. It is shown that the medium with the largest mass transfer resistance need only be considered to provide a conservative estimate of mass transfer rates. (author) 11 refs., 3 figs

Sampling and analysis validates acceptable knowledge on LANL transuranic, heterogeneous, debris waste, or ''Cutting the Gordian knot that binds WIPP''

International Nuclear Information System (INIS)

Kosiewicz, S.T.; Triay, I.R.; Souza, L.A.

1999-01-01

Through sampling and toxicity characteristic leaching procedure (TCLP) analyses, LANL and the DOE validated that a LANL transuranic (TRU) waste (TA-55-43, Lot No. 01) was not a Resource Recovery and Conservation Act (RCRA) hazardous waste. This paper describes the sampling and analysis project as well as the statistical assessment of the analytical results. The analyses were conducted according to the requirements and procedures in the sampling and analysis plan approved by the New Mexico Environmental Department. The plan used a statistical approach that was consistent with the stratified, random sampling requirements of SW-846. LANL adhered to the plan during sampling and chemical analysis of randomly selected items of the five major types of materials in this heterogeneous, radioactive, debris waste. To generate portions of the plan, LANL analyzed a number of non-radioactive items that were representative of the mix of items present in the waste stream. Data from these cold surrogates were used to generate means and variances needed to optimize the design. Based on statistical arguments alone, only two samples from the entire waste stream were deemed necessary, however a decision was made to analyze at least two samples of each of the five major waste types. To obtain these samples, nine TRU waste drums were opened. Sixty-six radioactively contaminated and four non-radioactive grab samples were collected. Portions of the samples were composited for chemical analyses. In addition, a radioactively contaminated sample of rust-colored powder of interest to the New Mexico Environment Department (NMED) was collected and qualitatively identified as rust

Waste retrieval sluicing system campaign number 1 solids volume transferred calculation

International Nuclear Information System (INIS)

BAILEY, J.W.

1999-01-01

This calculation has been prepared to document the volume of sludge removed from tank 241-C-106 during Waste Retrieval Sluicing System (WRSS) Sluicing Campaign No.1. This calculation will be updated, if necessary, to incorporate new data. This calculation supports the declaration of completion of WRSS Campaign No.1 and, as such, is also the documentation for completion of Performance Agreement TWR 1.2.1 , C-106 Sluicing Performance Expectations. It documents the performance of all the appropriate tank 241-C-106 mass transfer verifications, evaluations, and appropriate adjustments discussed in HNF-SD-WM-PROC-021, Chapter 23, ''Process Engineering Calculations for Tank 241-C-106 Sluicing and Retrieval''

Waste retrieval sluicing system campaign number 1 solids volume transferred calculation

International Nuclear Information System (INIS)

BAILEY, J.W.

1999-01-01

This calculation has been prepared to document the volume of sludge removed from tank 241-C-106 during Waste Retrieval Sluicing System (WRSS) Sluicing Campaign No.1. This calculation will be updated, if necessary, to incorporate new data. This calculation supports the declaration of completion of WRSS Campaign No.1 and, as such, is also the documentation for completion of Performance Agreement TWR 1.2.1 C-106 Sluicing Performance Expectations. It documents the performance of all the appropriate tank 241-C-106 mass transfer verifications, evaluations, and appropriate adjustments discussed in HNF-SD-WM-PROC-021, Chapter 23, ''Process Engineering Calculations for Tank 241-C-106 Sluicing and Retrieval''

Mission analysis for cross-site transfer

International Nuclear Information System (INIS)

Riesenweber, S.D.; Fritz, R.L.; Shipley, L.E.

1995-11-01

The Mission Analysis Report describes the requirements and constraints associated with the Transfer Waste Function as necessary to support the Manage Tank Waste, Retrieve Waste, and Process Tank Waste Functions described in WHC-SD-WM-FRD-020, Tank Waste Remediation System (TWRS) Functions and Requirements Document and DOE/RL-92-60, Revision 1, TWRS Functions and Requirements Document, March 1994. It further assesses the ability of the ''initial state'' (or current cross-site transfer system) to meet the requirements and constraints

Waste to wealth: Production of oxytetracycline using streptomyces ...

African Journals Online (AJOL)

Yomi

2012-05-29

May 29, 2012 ... Chemical structure: Sulfonamides and related drugs; diaminopyrimidines .... tetracycline is excreted in urine by glomerular filtration. They are partly .... paper. The filtrate was then transferred and stored in a sterile ... samples of both species of cocoyam peels wastes gave .... environmental problem. Thus, the ...

The International Stripa Project: Technology transfer from cooperation in scientific and technological research on nuclear waste disposal

International Nuclear Information System (INIS)

Levich, R.A.; Ferrigan, P.M.; Wilkey, P.L.

1990-01-01

The Nuclear Energy Agency of the organization for Economic Cooperation and Development (OECD/NEA) sponsors the International Stripa Project. The objectives of the Stripa Project are to develop techniques for characterizing sites located deep in rock formations that are potentially suitable for the geologic disposal of high-level radioactive wastes and to evaluate particular engineering design considerations that could enhance the long-term safety of a high-level radioactive waste repository in a geologic medium. The purpose of this paper is to briefly summarize the research conducted at Stripa and discuss the ways in which the technology developed for the Stripa Project has been and will be transfered to the United States Civilian Radioactive Waste Management Program's Yucca Mountain Project. 3 refs., 2 figs

Shielded cells transfer automation

International Nuclear Information System (INIS)

Fisher, J.J.

1984-01-01

Nuclear waste from shielded cells is removed, packaged, and transferred manually in many nuclear facilities. Radiation exposure is absorbed by operators during these operations and limited only through procedural controls. Technological advances in automation using robotics have allowed a production waste removal operation to be automated to reduce radiation exposure. The robotic system bags waste containers out of glove box and transfers them to a shielded container. Operators control the system outside the system work area via television cameras. 9 figures

Geochemical sampling scheme optimization on mine wastes based on hyperspectral data

CSIR Research Space (South Africa)

Zhao, T

2008-07-01

Full Text Available decontamination, for example, acid-generating minerals. Acid rock drainage can adversely have an impact on the quality of drinking water and the health of riparian ecosystems. To assess or monitor environmental impact of mining, sampling of mine waste is required...

Hanford high level waste: Sample Exchange/Evaluation (SEE) Program

International Nuclear Information System (INIS)

King, A.G.

1994-08-01

The Pacific Northwest Laboratory (PNL)/Analytical Chemistry Laboratory (ACL) and the Westinghouse Hanford Company (WHC)/Process Analytical Laboratory (PAL) provide analytical support services to various environmental restoration and waste management projects/programs at Hanford. In response to a US Department of Energy -- Richland Field Office (DOE-RL) audit, which questioned the comparability of analytical methods employed at each laboratory, the Sample Exchange/Exchange (SEE) program was initiated. The SEE Program is a selfassessment program designed to compare analytical methods of the PAL and ACL laboratories using sitespecific waste material. The SEE program is managed by a collaborative, the Quality Assurance Triad (Triad). Triad membership is made up of representatives from the WHC/PAL, PNL/ACL, and WHC Hanford Analytical Services Management (HASM) organizations. The Triad works together to design/evaluate/implement each phase of the SEE Program

Linear model correction: A method for transferring a near-infrared multivariate calibration model without standard samples

Science.gov (United States)

Liu, Yan; Cai, Wensheng; Shao, Xueguang

2016-12-01

Calibration transfer is essential for practical applications of near infrared (NIR) spectroscopy because the measurements of the spectra may be performed on different instruments and the difference between the instruments must be corrected. For most of calibration transfer methods, standard samples are necessary to construct the transfer model using the spectra of the samples measured on two instruments, named as master and slave instrument, respectively. In this work, a method named as linear model correction (LMC) is proposed for calibration transfer without standard samples. The method is based on the fact that, for the samples with similar physical and chemical properties, the spectra measured on different instruments are linearly correlated. The fact makes the coefficients of the linear models constructed by the spectra measured on different instruments are similar in profile. Therefore, by using the constrained optimization method, the coefficients of the master model can be transferred into that of the slave model with a few spectra measured on slave instrument. Two NIR datasets of corn and plant leaf samples measured with different instruments are used to test the performance of the method. The results show that, for both the datasets, the spectra can be correctly predicted using the transferred partial least squares (PLS) models. Because standard samples are not necessary in the method, it may be more useful in practical uses.

Parameters, test criteria and fault assessment in random sampling of waste barrels from non-qualified processes

International Nuclear Information System (INIS)

Martens, B.R.

1989-01-01

In the context of random sampling tests, parameters are checked on the waste barrels and criteria are given on which these tests are based. Also, it is shown how faulty data on the properties of the waste or faulty waste barrels should be treated. To decide the extent of testing, the properties of the waste relevant to final storage are determined based on the conditioning process used. (DG) [de

Efficient sample preparation method based on solvent-assisted dispersive solid-phase extraction for the trace detection of butachlor in urine and waste water samples.

Science.gov (United States)

Aladaghlo, Zolfaghar; Fakhari, Alireza; Behbahani, Mohammad

2016-10-01

In this work, an efficient sample preparation method termed solvent-assisted dispersive solid-phase extraction was applied. The used sample preparation method was based on the dispersion of the sorbent (benzophenone) into the aqueous sample to maximize the interaction surface. In this approach, the dispersion of the sorbent at a very low milligram level was achieved by inserting a solution of the sorbent and disperser solvent into the aqueous sample. The cloudy solution created from the dispersion of the sorbent in the bulk aqueous sample. After pre-concentration of the butachlor, the cloudy solution was centrifuged and butachlor in the sediment phase dissolved in ethanol and determined by gas chromatography with flame ionization detection. Under the optimized conditions (solution pH = 7.0, sorbent: benzophenone, 2%, disperser solvent: ethanol, 500 μL, centrifuged at 4000 rpm for 3 min), the method detection limit for butachlor was 2, 3 and 3 μg/L for distilled water, waste water, and urine sample, respectively. Furthermore, the preconcentration factor was 198.8, 175.0, and 174.2 in distilled water, waste water, and urine sample, respectively. Solvent-assisted dispersive solid-phase extraction was successfully used for the trace monitoring of butachlor in urine and waste water samples. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Sampling and Analysis Plan for Tank 241-AP-108 Waste in Support of Evaporator Campaign 2000-1

International Nuclear Information System (INIS)

RASMUSSEN, J.H.

2000-01-01

This Tank Sampling and Analysis Plan (TSAP) identifies sample collection, laboratory analysis, quality assurance/quality control (QA/QC), and reporting objectives for the characterization of tank 241-AP-108 waste. Technical bases for these objectives are specified in the 242-A Evaporator Data Quality Objectives (Bowman 2000 and Von Bargen 1998) and 108-AP Tank Sampling Requirements in Support of Evaporator Campaign 2000-1 (Le 2000). Evaporator campaign 2000-1 will process waste from tank 241-AP-108 in addition to that from tank 241-AP-107. Characterization results will be used to support the evaporator campaign currently planned for early fiscal year 2000. No other needs (or issues) requiring data for this tank waste apply to this sampling event

Resolving legal issues in the transfer of technologies in environmental restoration and waste management

International Nuclear Information System (INIS)

Katz, J.; Richards, F. III; Underwood, J.P.

1994-01-01

In reforming its contracting and procurement processes, the Department of Energy (DOE) should also make reforms to facilitate the transfer and commercialization of environmental restoration and waste management technologies between DOE laboratories and the private sector. These reforms would address the three related legal issues of patents, conflict of interest, and liability. This paper discusses each issue and considers possible solutions to them. These solutions include contractual clauses to specifically address the needs of all involved parties, improved definition of development and implementation rights, and increased protections for contractors. If DOE, or the Federal Government more generally, successfully resolves these issues and makes the appropriate changes to the DOE procurement system, transfer and commercialization will more efficiently, effectively, and easily occur

Wipe sampling for characterization of noncompactable radioactive waste

International Nuclear Information System (INIS)

Barbieri, Aline E.O.; Ferreira, Robson J.; Vicente, Roberto

2009-01-01

Wipe sampling is a method of monitoring radioactive surface contamination on working area and on radioactive, non-compactable wastes, constituted of large pieces of replaced parts of equipment in nuclear and radioactive installations. In this method, sampling is executed by rubbing a disc of filter paper on the contaminated surface in such a way as to collect entirely or partially the deposited material. The target radioisotopes are subsequently measured directly on the wipe or extracted by appropriate radio analytical methods and then qualitatively and quantitatively determined. The collection factor, or the efficiency with which the material is removed from the surface and deposited on the smear, is the main source of error in quantitative measurements. The determination of the collection efficiency is the object of this communication. (author)

Rapid analysis of mixed waste samples via the optical emission from laser initiated microplasmas

International Nuclear Information System (INIS)

Barefield, J.E. II; Ferran, M.D.; Cremers, D.A.

1993-01-01

Wavelength resolved optical emission from laser initiated microplasmas in samples containing Pu, Am, Pb, Cr, and Be was used to determine elemental compositions. Traditionally, samples of this type are analyzed by neutron activation, X-ray fluorescence, atomic absorption (AA), inductively coupled plasma - atomic emission spectroscopy (ICP-AES), and inductively coupled plasma mass spectroscopy (ICP-MS). Analysis via the traditional analytical spectroscopic techniques involves extensive sample separation and preparation which results in the generation of significant quantities of additional waste. In the laser based method, little to no sample preparation is required. The method is essentially waste free since only a few micrograms of material is removed from the sample in the generation of the microplasma. Detection limits of the laser based method typically range between subppm to tens of ppM. In this report, the optical emission from samples containing Pu, Am, Pb, Cr, and Be will be discussed. we will also discuss the essential elements of the analysis method

Replacement of the cross-site transfer system liquid waste transport alternatives evaluation, Project W-058

International Nuclear Information System (INIS)

Vo, D.V.; Epperson, E.M.

1995-05-01

This document examines high-/low-level radioactive liquid waste transport alternatives. Radioactive liquid waste will be transported from the 200 West Area to the 200 East Area and within the 200 East Areas for safe storage and disposal. The radioactive waste transport alternatives are the Aboveground Transport System (French LR-56 Cask System [3,800 L (1,000 gal)]), 19,000-L (5,000-gal) trailer tanker system, 75,700-L (20,000-gal) rail tanker system and Underground Transport System (buried pipe [unlimited transfer volume capability]). The evaluation focused on the following areas: initial project cost, operational cost, secondary waste generation, radiation exposure, and final decommissioning. The evaluation was based on the near term (1995 to 2005) estimated volume of 49.509 million L (13.063 million gal) and long term (1995 to 2028) estimated volume of 757.1 million L (200 million gal). The conclusion showed that the buried pipe (Underground Transport System) resulted in the lowest overall total cost for near and long term, the trailer container resulted in the highest total cost for near and long term, and the French truck was operationally impractical and cost prohibitive

Joint Coordinating Committee on environmental restoration and waste management (JCCEM) support, technology transfer, and special projects

International Nuclear Information System (INIS)

Edgar, D.E.

1993-01-01

Argonne National Laboratory (ANL) assisted in identifying and evaluating foreign technologies to meet EM needs; supported the evaluation, removal, and/or revision of barriers to international technology and information transfer/exchange; facilitated the integration and coordination of U.S. government international environmental restoration and waste management activities; and enhanced U.S. industry's competitiveness in the international environmental technology market

Data Quality Objectives for Regulatory Requirements for Dangerous Waste Sampling and Analysis; FINAL

International Nuclear Information System (INIS)

MULKEY, C.H.

1999-01-01

This document describes sampling and analytical requirements needed to meet state and federal regulations for dangerous waste (DW). The River Protection Project (RPP) is assigned to the task of storage and interim treatment of hazardous waste. Any final treatment or disposal operations, as well as requirements under the land disposal restrictions (LDRs), fall in the jurisdiction of another Hanford organization and are not part of this scope. The requirements for this Data Quality Objective (DQO) Process were developed using the RPP Data Quality Objective Procedure (Banning 1996), which is based on the U.S. Environmental Protection Agency's (EPA) Guidance for the Data Quality Objectives Process (EPA 1994). Hereafter, this document is referred to as the DW DQO. Federal and state laws and regulations pertaining to waste contain requirements that are dependent upon the composition of the waste stream. These regulatory drivers require that pertinent information be obtained. For many requirements, documented process knowledge of a waste composition can be used instead of analytical data to characterize or designate a waste. When process knowledge alone is used to characterize a waste, it is a best management practice to validate the information with analytical measurements

Characteristics of soil-to-plant transfer of elements relevant to radioactive waste in boreal forest

Energy Technology Data Exchange (ETDEWEB)

Roivainen, P.

2011-07-01

The use of nuclear energy generates large amounts of different types of radioactive wastes that can be accidentally released into the environment. Soil-to-plant transfer is a key process for the dispersion of radionuclides in the biosphere and is usually described by a concentration ratio (CR) between plant and soil concentrations in radioecological models. Our knowledge of the soil-to-plant transfer of many radionuclides is currently limited and concerns mainly agricultural species and temperate environments. The validity of radioecological modelling is affected by the accuracy of the assumptions and parameters used to describe soil-to-plant transfer. This study investigated the soil-to-plant transfer of six elements (cobalt (Co), molybdenum (Mo), nickel (Ni), lead (Pb), uranium (U) and zinc (Zn)) relevant to radioactive waste at two boreal forest sites and assessed the factors affecting the CR values. May lily (Maianthemum bifolium), narrow buckler fern (Dryopteris carthusiana) and blueberry (Vaccinium myrtillus) were selected as representatives of understory species, while rowan (Sorbus aucuparia) and Norway spruce (Picea abies) represented trees in this study. All the elements studied were found to accumulate in plant roots, indicating that separate CR values for root and aboveground plant parts are needed. The between-species variation in CR values was not clearly higher than the within-species variation, suggesting that the use of generic CR values for understory species and trees is justified. No linear relationship was found between soil and plant concentrations for the elements studied and a non-linear equation was found to be the best for describing the dependence of CR values on soil concentration. Thus, the commonly used assumption of a linear relationship between plant and soil concentrations may lead to underestimation of plant root uptake at low soil concentrations. Plant nutrients potassium, magnesium, manganese, phosphorus and sulphur were found to

Characteristics of soil-to-plant transfer of elements relevant to radioactive waste in boreal forest

International Nuclear Information System (INIS)

Roivainen, P.

2011-01-01

The use of nuclear energy generates large amounts of different types of radioactive wastes that can be accidentally released into the environment. Soil-to-plant transfer is a key process for the dispersion of radionuclides in the biosphere and is usually described by a concentration ratio (CR) between plant and soil concentrations in radioecological models. Our knowledge of the soil-to-plant transfer of many radionuclides is currently limited and concerns mainly agricultural species and temperate environments. The validity of radioecological modelling is affected by the accuracy of the assumptions and parameters used to describe soil-to-plant transfer. This study investigated the soil-to-plant transfer of six elements (cobalt (Co), molybdenum (Mo), nickel (Ni), lead (Pb), uranium (U) and zinc (Zn)) relevant to radioactive waste at two boreal forest sites and assessed the factors affecting the CR values. May lily (Maianthemum bifolium), narrow buckler fern (Dryopteris carthusiana) and blueberry (Vaccinium myrtillus) were selected as representatives of understory species, while rowan (Sorbus aucuparia) and Norway spruce (Picea abies) represented trees in this study. All the elements studied were found to accumulate in plant roots, indicating that separate CR values for root and aboveground plant parts are needed. The between-species variation in CR values was not clearly higher than the within-species variation, suggesting that the use of generic CR values for understory species and trees is justified. No linear relationship was found between soil and plant concentrations for the elements studied and a non-linear equation was found to be the best for describing the dependence of CR values on soil concentration. Thus, the commonly used assumption of a linear relationship between plant and soil concentrations may lead to underestimation of plant root uptake at low soil concentrations. Plant nutrients potassium, magnesium, manganese, phosphorus and sulphur were found to

Development of Simulants to Support Mixing Tests for High Level Waste and Low Activity Waste

International Nuclear Information System (INIS)

EIBLING, RUSSELLE.

2004-01-01

The objectives of this study were to develop two different types of simulants to support vendor agitator design studies and mixing studies. The initial simulant development task was to develop rheologically-bounding physical simulants and the final portion was to develop a nominal chemical simulant which is designed to match, as closely as possible, the actual sludge from a tank. The physical simulants to be developed included a lower and upper rheologically bounded: pretreated low activity waste (LAW) physical simulant; LAW melter feed physical simulant; pretreated high level waste (HLW) physical simulant; HLW melter feed physical simulant. The nominal chemical simulant, hereafter referred to as the HLW Precipitated Hydroxide simulant, is designed to represent the chemical/physical composition of the actual washed and leached sludge sample. The objective was to produce a simulant which matches not only the chemical composition but also the physical properties of the actual waste sample. The HLW Precipitated Hydroxide simulant could then be used for mixing tests to validate mixing, homogeneity and representative sampling and transferring issues. The HLW Precipitated Hydroxide simulant may also be used for integrated nonradioactive testing of the WTP prior to radioactive operation

REMOTE IN-CELL SAMPLING IMPROVEMENTS PROGRAM AT THESAVANNAH RIVER SITE (SRS) DEFENSE WASTE PROCESSING FACILITY (DWPF)

International Nuclear Information System (INIS)

Marzolf, A

2007-01-01

Remote Systems Engineering (RSE) of the Savannah River National Lab (SRNL) in combination with the Defense Waste Processing Facility(DWPF) Engineering and Operations has evaluated the existing equipment and processes used in the facility sample cells for 'pulling' samples from the radioactive waste stream and performing equipment in-cell repairs/replacements. RSE has designed and tested equipment for improving remote in-cell sampling evolutions and reducing the time required for in-cell maintenance of existing equipment. The equipment within the present process tank sampling system has been in constant use since the facility start-up over 17 years ago. At present, the method for taking samples within the sample cells produces excessive maintenance and downtime due to frequent failures relative to the sampling station equipment and manipulator. Location and orientation of many sampling stations within the sample cells is not conducive to manipulator operation. The overextension of manipulators required to perform many in-cell operations is a major cause of manipulator failures. To improve sampling operations and reduce downtime due to equipment maintenance, a Portable Sampling Station (PSS), wireless in-cell cameras, and new commercially available sampling technology has been designed, developed and/or adapted and tested. The uniqueness of the design(s), the results of the scoping tests, and the benefits relative to in-cell operation and reduction of waste are presented

Tank Farm Waste Transfer Compatibility Program

International Nuclear Information System (INIS)

FOWLER, K.D.

2001-01-01

The compatibility program described in this document formalizes the process for determining waste compatibility. The primary goal of the program is to ensure that sufficient controls are in place to prevent the formation of incompatible mixtures during future operations. The process described involves characterizing waste, comparing characteristics with criteria, resolving potential incompatibilities and documenting the process

Tank Farm Waste Transfer Compatibility Program

International Nuclear Information System (INIS)

FOWLER, K.D.

2000-01-01

The compatibility program described in this document formalizes the process for determining waste compatibility. The primary goal of the program is to ensure that sufficient controls are in place to prevent the formation of incompatible mixtures during future operations. The process described involves characterizing waste, comparing characteristics with criteria, resolving potential incompatibilities and documenting the process

Experimental study on heat transfer performance of fin-tube exchanger and PSHE for waste heat recovery

Science.gov (United States)

Chen, Ting; Bae, Kyung Jin; Kwon, Oh Kyung

2018-02-01

In this paper, heat transfer characteristics of fin-tube heat exchanger and primary surface heat exchanger (PSHE) used in waste heat recovery were investigated experimentally. The flow in the fin-tube heat exchanger is cross flow and in PSHE counter flow. The variations of friction factor and Colburn j factor with air mass flow rate, and Nu number with Re number are presented. Various comparison methods are used to evaluate heat transfer performance, and the results show that the heat transfer rate of the PSHE is on average 17.3% larger than that of fin-tube heat exchanger when air mass flow rate is ranging from 1.24 to 3.45 kg/min. However, the PSHE causes higher pressure drop, and the fin-tube heat exchanger has a wider application range which leads to a 31.7% higher value of maximum heat transfer rate compared to that of the PSHE. Besides, under the same fan power per unit frontal surface, a higher heat transfer rate value is given in the fin-tube heat exchanger.

Quantitative and Qualitative Investigation of Yazd Dental Center Waste

Directory of Open Access Journals (Sweden)

Mehdi Mokhtari

2015-12-01

Full Text Available Compared to other medical clinics, dental centers produce a relatively small amount of healthcare waste. However, they are responsible for the production of certain amounts of waste that can cause serious health and environmental hazards if not treated properly. Determining the quantity and quality of dental waste is quite necessary to plan the machinery, personnel, and transportation methods. In the current analytical cross-sectional study, 22 dental clinics were randomly selected from among 145 dental centers in Yazd. For each clinic, three samples were manually collected and distributed at the end of the work day once a week (the day was randomly chosen. The components were classified into four groups based on their characteristics and potential risk. Afterward, the collected data were analyzed using Excel. The mean per capita daily waste generation in the Yazd dental clinics was 80.179 g for each patient. In general, the Yazd dental clinics generate approximately 4 tons of waste every year. Out of this amount, infectious waste, domestic waste, chemical-pharmaceutical waste, and sharp waste account for 49.30%, 33.33%, 13.7%, and 4.2%, respectively. The results of the current study and the small amount of waste generated in Yazd suggest that a special safe site should be constructed for infectious, sharp, and keen waste; private companies should collect such waste and transfer it to this site after the waste is disinfected; and the waste should then be transferred to a municipal waste hygienic disposal site. In regard to chemical waste that may contain hazardous materials, like mercury, no suggestions other than safe burial are made.

Municipal solid waste composition determination supporting the integrated solid waste management system in the island of Crete

International Nuclear Information System (INIS)

Gidarakos, E.; Havas, G.; Ntzamilis, P.

2006-01-01

A one-year survey was conducted in the greater region of Crete (located at the lower region of the Aegean Sea) for the purpose of identifying waste composition (including chemical and physical characterization), as well as any seasonal variation. The investigation was carried out repeatedly at seven landfills and one transfer station in Crete, in four phases. Each sampling phase corresponded to a season (autumn, winter, spring, summer). ASTM D5231-92(2003) standard method and RCRA Waste Sampling Draft Technical Guidance were used. Hand sorting was used for classifying the collected wastes into the following categories: plastics, paper, metals, aluminium, leather-wood-textiles-rubbers, organic wastes, non-combustibles and miscellaneous. Further analysis included proximate and ultimate analysis of combustible materials. Metals such as lead, cadmium and mercury were also investigated. The results show that there has been a significant decrease of organic wastes during the last decade due to the increase of packaging materials, as a result of a change in consumption patterns. Three main waste categories were determined: organic wastes, paper and plastics, which combined represent 76% of the total waste in Crete. Furthermore, a high fraction of glass and a seasonal variation of aluminium indicate a strong correlation of waste composition with certain human activities, such as tourism. There is also a variation between the municipal solid waste (MSW) composition in the region of Crete (2003-2004) and MSW composition suggested in the National Solid Waste Planning (2000) [National Solid Waste Planning, 2000. Completion and particularization of Common Ministerial Act 113944//1944/1997: National Solid Waste Planning, June 2000]. The results of this survey are to be utilized by the regional solid waste authorities in order to establish an integrated waste treatment site, capable of fulfilling the regional waste management demands

Municipal solid waste composition determination supporting the integrated solid waste management system in the island of Crete.

Science.gov (United States)

Gidarakos, E; Havas, G; Ntzamilis, P

2006-01-01

A one-year survey was conducted in the greater region of Crete (located at the lower region of the Aegean Sea) for the purpose of identifying waste composition (including chemical and physical characterization), as well as any seasonal variation. The investigation was carried out repeatedly at seven landfills and one transfer station in Crete, in four phases. Each sampling phase corresponded to a season (autumn, winter, spring, summer). ASTM D5231-92(2003) standard method and RCRA Waste Sampling Draft Technical Guidance were used. Hand sorting was used for classifying the collected wastes into the following categories: plastics, paper, metals, aluminium, leather-wood-textiles-rubbers, organic wastes, non-combustibles and miscellaneous. Further analysis included proximate and ultimate analysis of combustible materials. Metals such as lead, cadmium and mercury were also investigated. The results show that there has been a significant decrease of organic wastes during the last decade due to the increase of packaging materials, as a result of a change in consumption patterns. Three main waste categories were determined: organic wastes, paper and plastics, which combined represent 76% of the total waste in Crete. Furthermore, a high fraction of glass and a seasonal variation of aluminium indicate a strong correlation of waste composition with certain human activities, such as tourism. There is also a variation between the municipal solid waste (MSW) composition in the region of Crete (2003-2004) and MSW composition suggested in the National Solid Waste Planning (2000) [National Solid Waste Planning, 2000. Completion and particularization of Common Ministerial Act 113944//1944/1997: National Solid Waste Planning, June 2000]. The results of this survey are to be utilized by the regional solid waste authorities in order to establish an integrated waste treatment site, capable of fulfilling the regional waste management demands.

Methods of sampling airborne fungi in working environments of waste treatment facilities

Czech Academy of Sciences Publication Activity Database

Černá, K.; Wittlingerová, Z.; Zimová, M.; Janovský, Zdeněk

2016-01-01

Roč. 29, č. 3 (2016), s. 493-502 ISSN 1232-1087 Institutional support: RVO:67985939 Keywords : airborne fungi * filter based bioaerosol sampling * waste sorting facility Subject RIV: EF - Botanics Impact factor: 0.930, year: 2016

Method of estimating maximum VOC concentration in void volume of vented waste drums using limited sampling data: Application in transuranic waste drums

International Nuclear Information System (INIS)

Liekhus, K.J.; Connolly, M.J.

1995-01-01

A test program has been conducted at the Idaho National Engineering Laboratory to demonstrate that the concentration of volatile organic compounds (VOCs) within the innermost layer of confinement in a vented waste drum can be estimated using a model incorporating diffusion and permeation transport principles as well as limited waste drum sampling data. The model consists of a series of material balance equations describing steady-state VOC transport from each distinct void volume in the drum. The primary model input is the measured drum headspace VOC concentration. Model parameters are determined or estimated based on available process knowledge. The model effectiveness in estimating VOC concentration in the headspace of the innermost layer of confinement was examined for vented waste drums containing different waste types and configurations. This paper summarizes the experimental measurements and model predictions in vented transuranic waste drums containing solidified sludges and solid waste

Characterizing the transformation and transfer of nitrogen during the aerobic treatment of organic wastes and digestates

Energy Technology Data Exchange (ETDEWEB)

Zeng Yang, E-mail: yang.zeng@irstea.fr [Irstea, UR GERE, 17 avenue de Cucille, CS 64427, F-35044 Rennes Cedex (France); Universite Europeenne de Bretagne, F-35000 Rennes (France); Guardia, Amaury de; Daumoin, Mylene; Benoist, Jean-Claude [Irstea, UR GERE, 17 avenue de Cucille, CS 64427, F-35044 Rennes Cedex (France)

2012-12-15

Highlights: Black-Right-Pointing-Pointer Ammonia emissions varied depending on the nature of wastes and the treatment conditions. Black-Right-Pointing-Pointer Nitrogen losses resulted from ammonia emissions and nitrification-denitrification. Black-Right-Pointing-Pointer Ammonification can be estimated from biodegradable carbon and carbon/nitrogen ratio. Black-Right-Pointing-Pointer Ammonification was the main process contributing to N losses. Black-Right-Pointing-Pointer Nitrification rate was negatively correlated to stripping rate of ammonia nitrogen. - Abstract: The transformation and transfer of nitrogen during the aerobic treatment of seven wastes were studied in ventilated air-tight 10-L reactors at 35 Degree-Sign C. Studied wastes included distinct types of organic wastes and their digestates. Ammonia emissions varied depending on the kind of waste and treatment conditions. These emissions accounted for 2-43% of the initial nitrogen. Total nitrogen losses, which resulted mainly from ammonia emissions and nitrification-denitrification, accounted for 1-76% of the initial nitrogen. Ammonification was the main process responsible for nitrogen losses. An equation which allows estimating the ammonification flow of each type of waste according to its biodegradable carbon and carbon/nitrogen ratio was proposed. As a consequence of the lower contribution of storage and leachate rates, stripping and nitrification rates of ammonia nitrogen were negatively correlated. This observation suggests the possibility of promotingnitrification in order to reduce ammonia emissions.

Submersible canned motor transfer pump

International Nuclear Information System (INIS)

Guardiani, R.F.; Pollick, R.D.; Nyilas, C.P.; Denmeade, T.J.

1997-01-01

A transfer pump is described which is used in a waste tank for transferring high-level radioactive liquid waste from a waste tank and having a column assembly, a canned electric motor means, and an impeller assembly with an upper impeller and a lower impeller connected to a shaft of a rotor assembly. The column assembly locates a motor housing with the electric motor means adjacent to the impeller assembly which creates an hydraulic head, and which forces the liquid waste, into the motor housing to cool the electric motor means and to cool and/or lubricate the radial and thrust bearing assemblies. Hard-on-hard bearing surfaces of the bearing assemblies and a ring assembly between the upper impeller and electric motor means grind large particles in the liquid waste flow. Slots in the static bearing member of the radial bearing assemblies further grind down the solid waste particles so that only particles smaller than the clearances in the system can pass there through, thereby resisting damage to and the interruption of the operation of the transfer pump. The column assembly is modular so that sections can be easily assembled, disassembled and/or removed. A second embodiment employs a stator jacket which provides an alternate means for cooling the electric motor means and lubricating and/or cooling the bearing assemblies, and a third embodiment employs a variable level suction device which allows liquid waste to be drawn into the transfer pump from varying and discrete levels in the waste tank. 17 figs

Design assessment for Melton Valley liquid low-level waste collection and transfer system upgrade project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

1994-10-01

This project is designed for collecting liquid low level waste (LLLW) from generating points inside the Radioisotope Engineering and Development Center (Buildings 7920 and 7930) facility and transferring this waste to the Collection Tank (F-1800) in the new Monitoring and Control Station (MCS) facility. The LLLW is transferred to the MCS in a new, underground, jacketed, stainless steel piping system. The LLLW will then be transferred from Tank F-1800 through a new, underground, jacketed, stainless steel piping system that connects the existing Bethel Valley LLLW Collection System and the Evaporator Facility Service Tanks. The interface for the two systems will be at the existing Interconnecting Pipe Line (ICPL) Valve Box adjacent to the Nonradiological Wastewater Treatment Plant. The project scope consists of the following systems: (1) Building 7920 LLLW Collection System; (2) Building 7930 LLLW Collection System; (3) LLLW Underground Transfer System to MCS; (4) MCS Building (including all equipment contained therein); (5) LLLW Underground Transfer System to ICPL Valve Box; and (6) Leak detection system for jacketed piping systems (3) and (5)

Mixed waste removal from a hazardous waste storage tank

International Nuclear Information System (INIS)

Geber, K.R.

1993-01-01

The spent fuel transfer canal at the Oak Ridge Graphite Reactor was found to be leaking 400 gallons of water per day into the surrounding soil. Sampling of the sediment layer on the floor of the canal to determine the environmental impact of the leak identified significant radiological contamination and elevated levels of cadmium and lead which are hazardous under the Resource Conservation and Recovery Act (RCRA). Under RCRA regulations and Rules of Tennessee Department of Environment and Conservation, the canal was considered a hazardous waste storage tank. This paper describes elements of the radiological control program established in support of a fast-track RCRA closure plan that involved underwater mapping of the radiation fields, vacuuming, and ultra-filtration techniques that were successfully used to remove the mixed waste sediments and close the canal in a method compliant with state and federal regulations

Waste compatibility assessments to support project W-320

International Nuclear Information System (INIS)

BLAAK, T.M.

1999-01-01

The intent of this internal memo is to provide a recommendation for the transfer of tank 241-C-106 waste, Attachment 2, to tank 241-AY-102. This internal memo also identifies additional requirements which have been deemed necessary for safely receiving and storing the waste documented in Attachment 2 from tank 241-C-106 in tank 241-AY-102. This waste transfer is planned in support of tank 241-C-106 solids sluicing activities. Approximately 200,000 gallons of waste and flush water are expected to be pumped from tank 241-C-106 into tank 241-AY-102. Several transfers will be necessary to complete the sluicing of tank 241-C-106 solids. To assure ourselves that this waste transfer will not create any compatibility concerns, a waste compatibility assessment adhering to current waste compatibility requirements has been performed

Analysis of Organic Samples from the 5-H and 3-F Pump Tanks and Waste Tank 38H

International Nuclear Information System (INIS)

Swingle, R.F. II

1999-01-01

Analyses for organic materials in aqueous and surface floating samples taken from the 5-H Pump Tank and Waste Tank 38H and in vapor samples taken from the 5-H and 3-F Pump Tanks have been completed. The results indicate that the concentration of organic materials is extremely low in all samples. This report documents the development of sampling and analysis techniques for this sampling as well as the results of the analyses of vapor samples pulled from Pump Tanks 5-H and 3-F and liquid samples pulled from Waste Tank 38H and Pump Tank 5-H

101-SY waste sample speed of sound/rheology testing for sonic probe program

International Nuclear Information System (INIS)

Cannon, N.S.

1994-01-01

One problem faced in the clean-up operation at Hanford is that a number of radioactive waste storage tanks are experiencing a periodic buildup and release of potentially explosive gases. The best known example is Tank 241-SY-101 (commonly referred to as 101-SY) in which hydrogen gas periodically built up within the waste to the point that increased buoyancy caused a roll-over event, in which the gas was suddenly released in potentially explosive concentrations (if an ignition source were present). The sonic probe concept is to generate acoustic vibrations in the 101-SY tank waste at nominally 100 Hz, with sufficient amplitude to cause the controlled release of hydrogen bubbles trapped in the waste. The sonic probe may provide a potentially cost-effective alternative to large mixer pumps now used for hydrogen mitigation purposes. Two important parameters needed to determine sonic probe effectiveness and design are the speed of sound and yield stress of the tank waste. Tests to determine these parameters in a 240 ml sample of 101-SY waste (obtained near the tank bottom) were performed, and the results are reported

Nonradioactive Dangerous Waste Landfill sampling and analysis plan and data quality objectives process summary report

International Nuclear Information System (INIS)

Smith, R.C.

1997-08-01

This sampling and analysis plan defines the sampling and analytical activities and associated procedures that will be used to support the Nonradioactive Dangerous Waste Landfill soil-gas investigation. This SAP consists of three sections: this introduction, the field sampling plan, and the quality assurance project plan. The field sampling plan defines the sampling and analytical methodologies to be performed

History of waste tank 16, 1959 through 1974

International Nuclear Information System (INIS)

Davis, T.L.; Tharin, D.W.; Jones, D.W.; Lohr, D.R.

1977-07-01

Tank 16 was placed in service as a receiver of fresh high heat waste (HW) on May 9, 1959, and was filled to capacity in May 1960. Approximately half the tank contents were transferred to tanks 14 and 15 during September and October 1960 because of leakage into the annulus. Use of tank 16 was resumed in October 1967 when authorization (TA 2-603) was obtained to receive LW, and the tank was filled to capacity by June 1968. Subsequently, supernate was removed from the tank, and a blend of fresh LW and evaporator bottoms was added. In March 1972, the supernate was transferred to tank 13 because leakage had resumed. The sludge was left in the tank bottom and the use of tank 16 for any additional waste storage was discontinued. In September 1960 liquid waste overflowed the annulus pan. Leakage essentially stopped after the tank liquid level was lowered below the middle horizontal weld. After exhaustive study, tank cracking and resultant leakage was concluded to have been caused by stress corrosion due to the action of NaOH or NaNO 3 on areas of high local stress in the steel plate such as welds. Samples of sludge, supernate, tank vapors, and leaked material in the annulus were analyzed, and tank temperature and radiation profiles were taken. Two disk samples were cut from the primary tank wall for metallurgical examination. Test coupons of various metals were exposed to tank 16 waste to aid new tank design and to study stress corrosion and hydrogen embrittlement. In addition, samples of SRP bedrock were placed in tank 16 to study reactions between bedrock and HW. 18 figures, 2 tables

Heat transfer: Pittsburgh 1987

International Nuclear Information System (INIS)

Lyczkowski, R.W.

1987-01-01

This book contains papers divided among the following sections: Process Heat Transfer; Thermal Hydraulics and Phase Change Phenomena; Analysis of Multicomponent Multiphase Flow and Heat Transfer; Heat Transfer in Advanced Reactors; General Heat Transfer in Solar Energy; Numerical Simulation of Multiphase Flow and Heat Transfer; High Temperature Heat Transfer; Heat Transfer Aspects of Severe Reactor Accidents; Hazardous Waste On-Site Disposal; and General Papers

Medicines discarded in household garbage: analysis of a pharmaceutical waste sample in Vienna.

Science.gov (United States)

Vogler, Sabine; Leopold, Christine; Zuidberg, Christel; Habl, Claudia

2014-01-01

To analyze a sample of pharmaceutical waste drawn from household garbage in Vienna, with the aim to learn whether and which medicines end up unused in normal household waste. We obtained a pharmaceutical waste sample from the Vienna Municipal Waste Department. This was drawn by their staff in a representative search in October and November 2009. We did a manual investigation of the sample which contained packs and loose blisters, excluded medical devices and traced loose blisters back to medicines packs. We reported information on the prescription status, origin, therapeutic group, dose form, contents and expiry date. We performed descriptive statistics for the total data set and for sub-groups (e.g. items still containing some of original content). In total, 152 packs were identified, of which the majority was prescription-only medicines (74%). Cardiovascular medicines accounted for the highest share (24%). 87% of the packs were in oral form. 95% of the packs had not expired. 14.5% of the total data set contained contents but the range of content left in the packs varied. Results on the packs with contents differed from the total: the shares of Over-the Counter medicines (36%), of medicines of the respiratory system (18%) and of the musculo-skeletal system (18%), for dermal use (23%) and of expired medicines (19%) were higher compared to the full data set. The study showed that some medicines end up unused or partially used in normal household garbage in Vienna. Our results did not confirm speculations about a high percentage of unused medicines improperly discarded. There is room for improved patient information and counseling to enhance medication adherence and a proper discharge of medicines.

Modeling transient heat transfer in nuclear waste repositories.

Science.gov (United States)

Yang, Shaw-Yang; Yeh, Hund-Der

2009-09-30

The heat of high-level nuclear waste may be generated and released from a canister at final disposal sites. The waste heat may affect the engineering properties of waste canisters, buffers, and backfill material in the emplacement tunnel and the host rock. This study addresses the problem of the heat generated from the waste canister and analyzes the heat distribution between the buffer and the host rock, which is considered as a radial two-layer heat flux problem. A conceptual model is first constructed for the heat conduction in a nuclear waste repository and then mathematical equations are formulated for modeling heat flow distribution at repository sites. The Laplace transforms are employed to develop a solution for the temperature distributions in the buffer and the host rock in the Laplace domain, which is numerically inverted to the time-domain solution using the modified Crump method. The transient temperature distributions for both the single- and multi-borehole cases are simulated in the hypothetical geological repositories of nuclear waste. The results show that the temperature distributions in the thermal field are significantly affected by the decay heat of the waste canister, the thermal properties of the buffer and the host rock, the disposal spacing, and the thickness of the host rock at a nuclear waste repository.

Mass transfer in a salt repository

International Nuclear Information System (INIS)

Pigford, T.H.; Chambre, P.L.

1985-05-01

To meet regulatory requirements for radioactive waste in a salt repository it is necessary to predict the rates of corrosion of the waste container, the release rates of radionuclides from the waste package, and the cumulative release of radionuclides into the accessible environment. The mechanisms that may control these rates and an approach to predicting these rates from mass-transfer theory are described. This new mechanistic approach is suggested by three premises: (a) a brine inclusion originally in a salt crystal moves along grain boundaries after thermal-induced migration out of the crystal, (b) brine moves along a grain boundary under the influence of a pressure gradient, and (c) salt surrounding a heat-generating waste package will soon creep and consolidate as a monolithic medium surrounding and in contact with the waste package. After consolidation there may be very little migration of intergranular and intragranular brine to the waste package. The corrosion rate of the waste container may then be limited by the rate at which brine reaches the container and may be calculable from mass-transfer theory, and the rate at which dissolved radionuclides leave the waste package may be limited by molecular diffusion in intragranular brine and may be calculable from mass-transfer theory. If porous nonsalt interbeds intersect the waste-package borehole, the release rate of dissolved radionuclides to interbed brine may also be calculable from mass-transfer theory. The logic of these conclusions is described, as an aid in formulating the calculations that are to be made

Characterization of hazardous waste sites: a methods manual. Volume 2. Available sampling methods (second edition)

International Nuclear Information System (INIS)

Ford, P.J.; Turina, P.J.; Seely, D.E.

1984-12-01

Investigations at hazardous waste sites and sites of chemical spills often require on-site measurements and sampling activities to assess the type and extent of contamination. This document is a compilation of sampling methods and materials suitable to address most needs that arise during routine waste site and hazardous spill investigations. The sampling methods presented in this document are compiled by media, and were selected on the basis of practicality, economics, representativeness, compatability with analytical considerations, and safety, as well as other criteria. In addition to sampling procedures, sample handling and shipping, chain-of-custody procedures, instrument certification, equipment fabrication, and equipment decontamination procedures are described. Sampling methods for soil, sludges, sediments, and bulk materials cover the solids medium. Ten methods are detailed for surface waters, groundwater and containerized liquids; twelve are presented for ambient air, soil gases and vapors, and headspace gases. A brief discussion of ionizing radiation survey instruments is also provided

Sampling and Analysis for Tank 241-AW-104 Waste in Support of Evaporator Campaign 2001-1

International Nuclear Information System (INIS)

MCKINNEY, S.G.

2000-01-01

This Tank Sampling and Analysis Plan (TSAP) identifies sample collection, laboratory analysis, quality assurance/quality control (QA/QC), and reporting objectives for the characterization of tank 241-AW-104 waste. Technical bases for these objectives are specified in the 242-A Evaporator Data Quality Objectives (Bowman 2000a and Von Bargen 1998), 242-A Evaporator Quality Assurance Project Plan (Bowman 1998 and Bowman 2000b), Tank 241-AW-104 Sampling Requirements in Support of Evaporator Campaign 2000-1 (Le 2000). Characterization results will be used to support the evaporator campaign currently planned for early fiscal year 2001. No other needs (or issues) requiring data for this tank waste apply to this sampling event

Estimates of laboratory accuracy and precision on Hanford waste tank samples

International Nuclear Information System (INIS)

Dodd, D.A.

1995-01-01

A review was performed on three sets of analyses generated in Battelle, Pacific Northwest Laboratories and three sets generated by Westinghouse Hanford Company, 222-S Analytical Laboratory. Laboratory accuracy and precision was estimated by analyte and is reported in tables. The sources used to generate this estimate is of limited size but does include the physical forms, liquid and solid, which are representative of samples from tanks to be characterized. This estimate was published as an aid to programs developing data quality objectives in which specified limits are established. Data resulting from routine analyses of waste matrices can be expected to be bounded by the precision and accuracy estimates of the tables. These tables do not preclude or discourage direct negotiations between program and laboratory personnel while establishing bounding conditions. Programmatic requirements different than those listed may be reliably met on specific measurements and matrices. It should be recognized, however, that these are specific to waste tank matrices and may not be indicative of performance on samples from other sources

ESP's Tank 42 washwater transfer to the 241-F/H tank farms

International Nuclear Information System (INIS)

Aponte, C.I.; Lee, E.D.

1997-12-01

As a result of the separation of the High-Level Liquid Waste Department into three separate organizations (formerly there were two) (Concentration, Storage, and Transfer (CST), Waste Pre-Treatment (WPT) and Waste Disposition (WD)) process interface controls were required. One of these controls is implementing the Waste the waste between CST and WPT. At present, CST's Waste Acceptance Criteria is undergoing revision and WPT has not prepared the required Waste Compliance Plan (WCP). The Waste Pre-Treatment organization is making preparations for transferring spent washwater in Tank 42 to Tank 43 and/or Tank 22. The washwater transfer is expected to complete the washing steps for preparing ESP batch 1B sludge. This report is intended to perform the function of a Waste Compliance Plan for the proposed transfer. Previously, transfers between the Tank Farm and ITP/ESP were controlled by requirements outlined in the Tank Farm's Technical Standards and ITP/ESP's Process Requirements. Additionally, these controls are implemented primarily in operating procedure 241-FH-7TSQ and ITP Operations Manual SW16.1-SOP-WTS-1 which will be completed prior to performing the waste transfers

Radioactive wastes handling facility

International Nuclear Information System (INIS)

Hirose, Emiko; Inaguma, Masahiko; Ozaki, Shigeru; Matsumoto, Kaname.

1997-01-01

There are disposed an area where a conveyor is disposed for separating miscellaneous radioactive solid wastes such as metals, on area for operators which is disposed in the direction vertical to the transferring direction of the conveyor, an area for receiving the radioactive wastes and placing them on the conveyor and an area for collecting the radioactive wastes transferred by the conveyor. Since an operator can conduct handling while wearing a working cloth attached to a partition wall as he wears his ordinary cloth, the operation condition can be improved and the efficiency for the separating work can be improved. When the area for settling conveyors and the area for the operators is depressurized, cruds on the surface of the wastes are not released to the outside and the working clothes can be prevented from being involved. Since the wastes are transferred by the conveyor, the operator's moving range is reduced, poisonous materials are fallen and moved through a sliding way to an area for collecting materials to be separated. Accordingly, the materials to be removed can be accumulated easily. (N.H.)

Study on sampling conditions for the monitoring of waste air

International Nuclear Information System (INIS)

Moeller, T.J.; Buetefisch, K.A.

1998-01-01

The technical codes for radiological monitoring of the waste air released from a radwaste repository demand that sampling for determination of aerosol-borne radioactivity is to be made with a screener equipped with a suitable number of measuring probes extending over the entire cross-sectional surface of the vent. Another requirement is to ensure that the waste air stream passing through the measuring channel is representative, containing the typical, operation-induced distribution of aerosols across the surface to be scanned. The study reported was intended to determine in a scaled-down model (1:10) of a repository ventilating duct the typical spatial distribution of aerosols (3D particulate density) in order to establish information on the type of typical distributions of aerosols, to be used for optimisation of the measuring site and monitoring instruments. (orig./CB) [de

Radiation protection in transference of radioactive wastes among buildings of an intermediary deposit

International Nuclear Information System (INIS)

Mitake, Malvina Boni; Suzuki, Fabio Fumio

2011-01-01

This paper describes the planning of radioprotection realized for transfer operation of radioactive wastes from two old buildings for a one of the new buildings. For planning purposes the operation was divided into nine stages and, for evaluation of collective dose, it was considered various relevant factors. The result of radioprotection optimization it was expected a total collective dose of 58.6 mSv per person. The measured dose per dosemeter of direct reading was of 3.9 mSv per person. These difference among the values is due to conservative factors used in the calculation

Long-term sampling of CO2 from waste-to-energy plants: 14C determination methodology, data variation and uncertainty

DEFF Research Database (Denmark)

Fuglsang, Karsten; Pedersen, Niels Hald; Larsen, Anna Warberg

2014-01-01

A dedicated sampling and measurement method was developed for long-term measurements of biogenic and fossil-derived CO2 from thermal waste-to-energy processes. Based on long-term sampling of CO2 and 14C determination, plant-specific emission factors can be determined more accurately, and the annual...... emission of fossil CO2 from waste-to-energy plants can be monitored according to carbon trading schemes and renewable energy certificates. Weekly and monthly measurements were performed at five Danish waste incinerators. Significant variations between fractions of biogenic CO2 emitted were observed...... was ± 4.0 pmC (95 % confidence interval) at 62 pmC. The long-term sampling method was found to be useful for waste incinerators for determination of annual fossil and biogenic CO2 emissions with relatively low uncertainty....

Identification of Heat Transfer Resistance of Scale Deposit on theEvaporator of Radioactive Waste Management Installation

International Nuclear Information System (INIS)

Zainus-Salimin

2000-01-01

Identification of heat transfer resistance of scale deposit from fixedhardness of liquid waste in the form of CaSO 4 and MgSO 4 ratio 2:1 has beendone on the evaporation system of Serpong Nuclear Facilities fordetermination of the quality of heat transfer obstruction from heating sourceto solution. Evaporation simulation of solution containing hardness withconcentration 0.5; 1; 2; and 2.5% mass were done on the stainless steelcontainer of 1 / volume with electrical heater in which a stainless-steeltube is put down on the base container. After 24, 168, 336, 504 and 672 hoursevaporation process it is obtained the thickness of scale deposit on thesurface of tube for determining the fouling factor. Heat transfer resistanceof scale deposit from 672 hours evaporation of solution 2.5% concentrationhampered heat transfer, the value of fouling factor be superior to limitsvalue of 0.000515 hours.m 2 . o C/kcal.The fouling factor from the evaporationof solution of 0.5; 1; and 2% concentration during 672 hours be inferior tolimits value. (author)

Identification of Cellulose Breaking Bacteria in Landfill Samples for Organic Waste Management

Science.gov (United States)

Chan, P. M.; Leung, F. C.

2015-12-01

According to the Hong Kong Environmental Protection Department, the citizens of Hong Kong disposes 13,500 tonnes of waste to the landfill everyday. Out of the 13,500 tonnes, 3600 tonnes consist of organic waste. Furthermore, due to the limited supply of land for landfills in Hong Kong, it is estimated that landfills will be full by about 2020. Currently, organic wastes at landfills undergo anaerobic respiration, where methane gas, one of the most harmful green house gases, will be released. The management of such waste is a pressing issue, as possible solutions must be presented in this crucial period of time. The Independent Schools Foundation Academy introduced their very own method to manage the waste produced by the students. With an approximate of 1500 students on campus, the school produces 27 metric tonnes of food waste each academic year. The installation of the rocket food composter provides an alternate method of disposable of organic waste the school produces, for the aerobic environment allows for different by-products to be produced, namely compost that can be used for organic farming by the primary school students and subsequently carbon dioxide, a less harmful greenhouse gas. This research is an extension on the current work, as another natural factor is considered. It evaluates the microorganism community present in leachate samples collected from the North East New Territories Landfill, for the bacteria in the area exhibits special characteristics in the process of decomposition. Through the sequencing and analysis of the genome of the bacteria, the identification of the bacteria might lead to a break through on the current issue. Some bacteria demonstrate the ability to degrade lignin cellulose, or assist in the production of methane gas in aerobic respirations. These characteristics can hopefully be utilized in the future in waste managements across the globe.

Design, placement, and sampling of groundwater monitoring wells for the management of hazardous waste disposal facilities

International Nuclear Information System (INIS)

Tsai, S.Y.

1988-01-01

Groundwater monitoring is an important technical requirement in managing hazardous waste disposal facilities. The purpose of monitoring is to assess whether and how a disposal facility is affecting the underlying groundwater system. This paper focuses on the regulatory and technical aspects of the design, placement, and sampling of groundwater monitoring wells for hazardous waste disposal facilities. Such facilities include surface impoundments, landfills, waste piles, and land treatment facilities. 8 refs., 4 figs

Sampling and analysis of high level waste tank supernatant: an overview

International Nuclear Information System (INIS)

Goergen, C.R.

1981-01-01

The Savannah River Plant routinely samples its high level radioactive waste tank supernatants for analysis of major components. These results are important in maintaining proper levels of corrosion inhibiters for protection of the tank walls. Because the tank ambient temperature is elevated, the sample is heated to 70 0 C prior to removing aliquots for use in a variety of analytical methods. Typical analyses include density, pH, OH - , NO 3 - , and NO 2 - , with occasional requests for Al(OH) 4 - , CO 3 /sup =/, PO 4 /sup =/, SO 4 /sup =/, and various radionuclides

Sample preparation of waste water to determine metallic contaminants by X-ray fluorescence spectroscopy

International Nuclear Information System (INIS)

Gonzalez Olivos, Javier.

1987-01-01

Trace X-ray fluorescence spectroscopy analysis in liquid samples is preceded by sample preparation, which usually consists in the precipitation of the metallic ions and concentration over a thin cellulose filter. The samples preparation of waste water by this method is not efficient, due to the great amount of organic and insoluble matter that they contain. The purpose of this work was to determine the optimal value of pH in order to adsorbe all the insoluble matter contained in a waste water sample in the activated charcoal, so that the metallic ions could be precipitated and concentrated on a thin filter and determinated by X-ray fluorescence spectroscopy. A survey about the adsorption of some ions in activated charcoal in function of the pH was made for the following: Cr 3+ , Fe 3+ , Ni 2+ , Cu 2+ , Zn 2+ , Se 2+ , Hg 2+ , and Pb 2+ . It was observed that at pH 0, the ions are not adsorbed, but Cu 2+ and Zn 2+ are adsorbed in small amount; at pH 14, the ions are adsorbed, excluding Se, which is not adsorbed at any value of pH. If a waste water sample is treated at pH 0 with activated charcoal to adsorbe the organic and insoluble matter, most of the metallic ions are not adsorbed by the activated charcoal and could be precipitated with APDC (ammonium 1-pirrolidine dithio carbamate salt) and concentrated on a thin filter. The analysis of the metallic ions contained on the filter and those adsorbed in the activated charcoal by X-ray fluorescence spectroscopy, gave the total amount of the ions in the sample. (author)

Fluorimetric determination of uranium in certain refractory minerals, environmental samples and industrial waste materials

International Nuclear Information System (INIS)

Premadas, A.; Saravanakumar, G.

2005-01-01

A simple sample decomposition and laser fluorimetric determination of uranium at trace level is reported in certain refractory minerals, like ilmenite, rutile, zircon and monazite; environmental samples viz. soil and sediments; industrial waste materials, such as, coal fly ash and red mud. Ilmenite sample is decomposed by heating with ammonium fluoride. Rutile, zircon and monazite minerals are decomposed by fusion using a mixture of potassium bifluoride and sodium fluoride. Environmental and industrial waste materials are brought into solution by treating with a mixture of hydrofluoric and nitric acids. The laser induced fluorimetric determination of uranium is carried out directly in rutile, zircon and in monazite minerals and after separation in other samples. The determination limit was 1 μg x g -1 for ilmenite, soil, sediment, coal fly ash and red mud samples, and it is 5 μg x g -1 for rutile, zircon and monazite. The method is also developed for the optical fluorimetric determination of uranium (determination limit 10 μg x g -1 ) in ilmenite, rutile, zircon and monazite minerals. The methods are simple, accurate, and precise and they require small quantity of sample and can be applied for the routine analysis. (author)

Components for containment enclosures - Part 3: Transfer systems such as plain doors, airlock chambers, double door transfer systems, leaktight connections for waste drums. 1. ed.

International Nuclear Information System (INIS)

1998-01-01

This part of ISO 11933 specifies requirements for the selection, construction and use of the following leak tight components: doors, airlock chambers, double door transfer systems, leaktight connections for waste drums. Some of the elements, double doors or airlock chambers are described in ISO 11933-1 and ISO 11933-2 as well. Doors having bigger dimensions used for personnel od larger objects are not covered by this document

Treatment and conditioning of historical radioactive waste

International Nuclear Information System (INIS)

Dogaru, Ghe.; Dragolici, F.; Ionascu, L.; Rotarescu, Ghe.

2009-01-01

The paper describes the management of historical radioactive waste from the storage facility of Radioactive Waste Treatment Plant. The historical waste stored into storage facility of IFIN-HH consists of spent sealed radioactive sources, empty contaminated containers, wooden radioactive waste, low specific activity radioactive waste, contaminated waste as well as radioactive waste from operation of WWR-S research reactor. After decommissioning of temporary storage facility about 5000 packages with radioactive waste were produced and transferred to the disposal facility. A large amount of packages have been transferred and disposed of to repository but at the end of 2000 there were still about 800 packages containing cement conditioned radioactive waste in an advanced state of degradation declared by authorities as 'historical waste'. During the management of historical waste campaign there were identified: radium spent radioactive sources, containers containing other spent sealed radioactive sources, packages containing low specific activity waste consist of thorium scrap allow, 30 larger packages (316 L), packages with activity lower than activity limit for disposal, packages with activity higher than activity limit for disposal. At the end of 2008, the whole amount of historical waste which met the waste acceptance criteria has been conditioned and transferred to disposal facility. (authors)

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 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.

Remote process cell mercury transfer pumps for DWPF

International Nuclear Information System (INIS)

Nielsen, M.G.; Vaughn, V.G.

1986-01-01

Final design and the results of the testing performed thus far show that the water displacement of mercury to a height of 40 feet is feasible with just 6 gallons of motive water. Control of the transfer is achieved by monitoring the pump discharge pressure. An air actuated plug valve configuration successfully contained the required discharge pressure of 260 psi. The requirements of low flow and maximum separation of mercury from particulates are achieved due to the configuration of the pressure canister. The pump is capable of transferring a discrete amount of mercury with little additional slurry particulates. The success of this new pumping configuration is highlighted by the fact that it was the inspiration for other remote transfer applications tested at SRP. These application include the dual canister sample pump shown in Figure 7, as well as a successful prototype pump designed at Pacific Northwest Laboratories (PNL). The PNL pump was designed for the purpose of metering waste slurries to an electric melter. Upon completion of final pump fabrication, the Defense Waste Processing facility (DWPF) facility will have a simple and highly reliable method of remotely transferring small discrete batches of mercury as required from radioactive process vessels. 3 refs., 7 figs., 1 tab

Basis for Selection of a Residual Waste Retrieval System for Gunite and Associated Tank W-9 at the Oak Ridge National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Lewis, B.E

2000-10-23

Waste retrieval and transfer operations at the Gunite{trademark} and Associated Tanks (GAATs) Remediation Project have been successfully accomplished using the Tank Waste Retrieval System. This system is composed of the Modified Light-Duty Utility Arm, Houdini Vehicle, Waste Dislodging and Conveyance System, Hose Management Arm, and Sludge Conditioning System. GAAT W-9 has been used as a waste-consolidation and batch-transfer tank during the retrieval of sludges and supernatants from the seven Gunite tanks in the North and South tank farms at Oak Ridge National Laboratory. Tank W-9 was used as a staging tank for the transfers to the Melton Valley Storage Tanks (MVSTs). A total of 18 waste transfers from W-9 occurred between May 25, 1999, and March 30, 2000. Most of these transfers were accomplished using the PulsAir Mixer to mobilize and mix the slurry and a submersible retrieval-transfer pump to transfer the slurry through the Sludge Conditioning System and the {approx}1-mile long, 2-in.-diam waste-transfer line to the MVSTs. The transfers from W-9 have consisted of low-solids-content slurries with solids contents ranging from {approx}2.8 to 6.8 mg/L. Of the initial {approx}88,000 gal of wet sludge estimated in the GAATs, a total of {approx}60,451 gal have been transferred to the MVSTs via tank W-9 as of March 30, 2000. Once the waste-consolidation operations and transfers from W-9 to the MVSTs are completed, the remaining material in W-9 will be mobilized and transferred to the active waste system, Bethel Valley Evaporator Service Tank W-23. Tank W-23 will serve as a batch tank for the final waste transfers from tank W-9 to the MVSTs. This report provides a summary of the requirements and recommendations for the final waste retrieval system for tank W-9, a compilation of the sample analysis data for the sludge in W-9, and brief descriptions of the various waste-retrieval system concepts that were considered for this task. The recommended residual waste retrieval

Double Shell Tank (DST) Transfer Pump Subsystem Specification

International Nuclear Information System (INIS)

GRAVES, C.E.

2001-01-01

This specification establishes the performance requirements and provides the references to the requisite codes and standards to be applied during the design of the Double-Shell Tank (DST) Transfer Pump Subsystem that supports the first phase of waste feed delivery (WFD). The DST Transfer Pump Subsystem consists of a pump for supernatant and/or slurry transfer for the DSTs that will be retrieved during the Phase 1 WFD operations. This system is used to transfer low-activity waste (LAW) and high-level waste (HLW) to designated DST staging tanks. It also will deliver blended LAW and HLW feed from these staging tanks to the River Protection Project (RPP) Waste Treatment Plant where it will be processed into an immobilized waste form. This specification is intended to be the basis for new projects/installations (W-521, etc.). This specification is not intended to retroactively affect previously established project design criteria without specific direction by the program

Preventing Buoyant Displacement Gas Release Events in Hanford Double-Shell Waste Tanks

Energy Technology Data Exchange (ETDEWEB)

Meyer, Perry A.; Stewart, Charles W.

2001-01-01

This report summarizes the predictive methods used to ensure that waste transfer operations in Hanford waste tanks do not create waste configurations that lead to unsafe gas release events. The gas release behavior of the waste in existing double-shell tanks has been well characterized, and the flammable gas safety issues associated with safe storage of waste in the current configuration are being formally resolved. However, waste is also being transferred between double-shell tanks and from single-shell tanks into double-shell tanks by saltwell pumping and sluicing that create new wastes and waste configurations that have not been studied as well. Additionally, planning is underway for various waste transfer scenarios to support waste feed delivery to the proposed vitrification plant. It is critical that such waste transfers do not create waste conditions with the potential for dangerous gas release events.

Treatment of mixed wastes by thermal plasma discharges

International Nuclear Information System (INIS)

Diaz A, L.V.

2007-01-01

The present study has as purpose to apply the technology of thermal plasma in the destruction of certain type of waste generated in the ININ. As first instance, origin, classification and disposition of the radioactive waste generated in the ININ is identified. Once identified the waste, the waste to treat is determined based on: the easiness of treating him with plasma, classification and importance. Later on, a substance or compound settles down (sample model) that serves as indicative of the waste for its physical-chemical characteristics, this is made because in the Thermal Plasma Applications Laboratory is not had the license to work with radioactive material. The sample model and the material to form the vitreous matrix are characterized before and after the treatment in order to evaluating their degradation and vitrification. During the treatment by means of the thermal plasma, the appropriate conditions are determined for the degradation and vitrification of the waste. Also, it is carried out an energy balance in the system to know the capacity to fuse the material depending the transfer of existent heat between the plasma and the material to treat. Obtaining favorable results, it thought about to climb in the project and by this way to help to solve one of the environmental problems in Mexico, as they are it the mixed wastes. (Author)

Removal of fission products from waste solutions using 16 different soil samples

International Nuclear Information System (INIS)

Bangash, M.A.; Hanif, J.

1997-01-01

Most of the nuclear sites use pits in the surrounding soils for the storage/disposal of low active waste (LAW) solutions. The characteristics of the soil if not suitable for the fixation or adsorption of the radioactive nuclides, may cause migration of these nuclides to hydrosphere. The phenomenon has the risk of radio toxic pollution for the living bodies therefore minerals composing the soil and their adsorption properties need to be investigated. For this purpose 16 different soil samples were collected from all over Pakistan. Mineralogical composition of the soils was determined by X-ray diffraction analysis. It was found that most of the samples contained clay minerals, illite, kaolinite and montmorillonite. Studies for the removal of fission products like, /sup 137/Cs. /sup 60/Sr and activation product /sup 60/CO from solution were carried out on these samples. The sorption experiments were performed by batch technique using radioactive as tracers. Distribution co-efficient were determined by mixing he element solution at pH 3 with the soil at soil solution ratios of 1 to 20. It is revealed from the experimental data that efficient removal of fission products from solutions is achieved by soil samples containing clay mineral montmorillonite, followed by little and kaolinite. These soils thus can be effectively used for the disposal of low level radioactive waste solutions without causing any environmental hazard. (author)

Preparation of radioactive ''mixed'' waste samples for measurement of RCRA [Resource Conservation and Recovery Act] organic compounds

International Nuclear Information System (INIS)

Tomkins, B.A.; Caton, J.E.

1987-01-01

A radioactive ''mixed'' waste typically contains alpha-, beta-, or gamma-emitting radionuclides and varying quantities of semivolatile or volatile organic species, some or all of which may be named specifically by the Resource Conservation and Recovery Act (RCRA). Because there are no acceptable means available currently for disposing of these mixed wastes, they are presently stored above-ground in sealed drums. For this reason, analytical procedures which can determine RCRA organics in radioactive waste are necessary for deciding the proper approach for disposal. An important goal of this work is the development of methods for preparing mixed waste samples in a manner which allows the RCRA organics to be measured in conventional organic analysis laboratories without special precautions. Analytical procedures developed for handling mixed waste samples must satisfy not only the usual constraints present in any trace-level organic chemical determination, but also those needed to insure the protection of the operator from radioactive contamination. Consequently, procedures should be designed to use the least amount of radioactive sample commensurate with achieving acceptable sensitivity with the RCRA analytical methods. Furthermore, the unusual laboratory glassware which would normally be used should be replaced with disposable materials wherever possible, in order to reduce the ''clean-up'' time required, and thereby reduce the operator's exposure to radioactivity. Actual sample handling should be reduced to the absolute minimum. Finally, the final isolate must exhibit a sufficiently low level of alpha, beta, or gamma activity to permit detailed characterization in a conventional organic analysis laboratory. 4 refs., 5 tabs

Preliminary evaluation of the impact and inter-generation risk transfers related to the release and disposal of radioactive waste from the nuclear fuel cycle

International Nuclear Information System (INIS)

Tort, V.; Lochard, J.; Schneider, T.; Sugier, A.

1997-12-01

This report is an attempt to contribute to the complex issue of the decision-making in the field of radioactive waste management. Because of the complex and multidimensional nature of the distant future consequences of waste management options, their analysis implies the taking into considerations of various aggregated indicators which depend on the elapse of time during which the radionuclides remain in the environment and their local, regional or world-wide dispersion. This report is a preliminary work sponsored by IPSN mainly focused on the risk transfer dimension, inherent to waste disposal management. Its objective is to illustrate, using the French nuclear fuel cycle context, the relative impact of some simple waste management options, outlining particularly the issue of inter-generation risk transfer. Even though the selected six radionuclides are the most important, a complete assessment should include all the radionuclides contained in the waste, what is particularly important in case of underground waste disposal were both normal evolution scenarios and intrusion must be considered. The extreme alternatives, i.e. the total disposal or total release of the radionuclides are analyzed but realistic are the intermediate options, which should be thoroughly examined from the technical point of view. The analysis of intermediate management options could give an estimation of the most appropriate solution in an ALARA perspective

The radioactivity estimation of 14C and 3H in graphite waste samples of the KRR-2.

Science.gov (United States)

Reyoung Kim, Hee

2013-09-01

The radioactivity of (14)C and (3)H in graphite samples from the dismantled Korea Research Reactor-2 (the KRR-2) site was analyzed by high-temperature oxidation and liquid scintillation counting, and the graphite waste was suggested to be disposed of as a low-level radioactive waste. The graphite samples were oxidized at a high temperature of 800 °C, and their counting rates were measured by using a liquid scintillation counter (LSC). The combustion ratio of the graphite was about 99% on the sample with a maximum weight of 1g. The recoveries from the combustion furnace were around 100% and 90% in (14)C and (3)H, respectively. The minimum detectable activity was 0.04-0.05 Bq/g for the (14)C and 0.13-0.15 Bq/g for the (3)H at the same background counting time. The activity of (14)C was higher than that of (3)H over all samples with the activity ratios of the (14)C to (3)H, (14)C/(3)H, being between 2.8 and 25. The dose calculation was carried out from its radioactivity analysis results. The dose estimation gave a higher annual dose than the domestic legal limit for a clearance. It was thought that the sampled graphite waste from the dismantled research reactor was not available for reuse or recycling and should be monitored as low-level radioactive waste. Copyright © 2013 Elsevier Ltd. All rights reserved.

West Valley waste removal system study

International Nuclear Information System (INIS)

Janicek, G.P.

1981-04-01

This study addresses the specific task of removing high-level wastes from underground tanks at Western New York Nuclear Center and delivering them to an onsite waste solidification plant. It begins with a review of the design and construction features of the waste storage tanks pertinent to the waste removal task with particular emphasis on the unique and complex tank internals which severely complicate the task of removal. It follows with a review of tank cleaning techniques used and under study at both Hanford and Savannah River and previous studies proposing the use of these techniques at West Valley. It concludes from these reviews that existing techniques are not directly transferable to West Valley and that a new approach is required utilizing selected feature and attributes from existing methodology. The study also concludes, from an investigation of the constraints imposed by the processing facility, that waste removal will be intermittent, requiring batch transfer over the anticipated 3 years of processing operations. Based on these reviews and conclusions, the study proposes that the acid waste be processed first and that one of the 15,000-gallon acid tanks then be used for batch feeding the neutralized waste. The proposed system would employ commercially available pumping equipment to transfer the wastes from the batch tank to processing via existing process piping. A commercially available mixed-flow pump and eight turbine pumps would homogenize the neutralized waste in conjunction with eight custom-fabricated sluicers for periodic transfer to the batch tank

Double Shell Tank (DST) Transfer Piping Subsystem Specification

International Nuclear Information System (INIS)

GRAVES, C.E.

2000-01-01

This specification establishes the performance requirements and provides references to the requisite codes and standards to be applied during design of the Double-Shell Tank (DST) Transfer Piping Subsystem that supports the first phase of Waste Feed Delivery. This specification establishes the performance requirements and provides references to the requisite codes and standards to be applied during design of the Double-Shell Tank (DST) Transfer Piping Subsystem that supports the first phase of waste feed delivery. This subsystem transfers waste between transfer-associated structures (pits) and to the River Protection Project (RPP) Privatization Contractor Facility where it will be processed into an immobilized waste form. This specification is intended to be the basis for new projects/installations (W-521, etc.). This specification is not intended to retroactively affect previously established project design criteria without specific direction by the program

Hanford Tank Waste Treatment and Immobilization Plant (WTP) Waste Feed Qualification Program Development Approach - 13114

Energy Technology Data Exchange (ETDEWEB)

Markillie, Jeffrey R.; Arakali, Aruna V.; Benson, Peter A.; Halverson, Thomas G. [Hanford Tank Waste Treatment and Immobilization Plant Project, Richland, WA 99354 (United States); Adamson, Duane J.; Herman, Connie C.; Peeler, David K. [Savannah River National Laboratory, Aiken, SC 29808 (United States)

2013-07-01

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is a nuclear waste treatment facility being designed and constructed for the U.S. Department of Energy by Bechtel National, Inc. and subcontractor URS Corporation (under contract DE-AC27-01RV14136 [1]) to process and vitrify radioactive waste that is currently stored in underground tanks at the Hanford Site. A wide range of planning is in progress to prepare for safe start-up, commissioning, and operation. The waste feed qualification program is being developed to protect the WTP design, safety basis, and technical basis by assuring acceptance requirements can be met before the transfer of waste. The WTP Project has partnered with Savannah River National Laboratory to develop the waste feed qualification program. The results of waste feed qualification activities will be implemented using a batch processing methodology, and will establish an acceptable range of operator controllable parameters needed to treat the staged waste. Waste feed qualification program development is being implemented in three separate phases. Phase 1 required identification of analytical methods and gaps. This activity has been completed, and provides the foundation for a technically defensible approach for waste feed qualification. Phase 2 of the program development is in progress. The activities in this phase include the closure of analytical methodology gaps identified during Phase 1, design and fabrication of laboratory-scale test apparatus, and determination of the waste feed qualification sample volume. Phase 3 will demonstrate waste feed qualification testing in support of Cold Commissioning. (authors)

SLUDGE BATCH 5 ACCEPTANCE EVALUATION RADIONUCLIDE CONCENTRATIONS IN TANK 51 SB5 QUALIFICATION SAMPLE PREPARED AT SRNL

International Nuclear Information System (INIS)

Bannochie, C; Ned Bibler, N; David Diprete, D

2008-01-01

Presented in this report are radionuclide concentrations required as part of the program of qualifying Sludge Batch Five (SB5) for processing in the Defense Waste Processing Facility (DWPF). Part of this SB5 material is currently in Tank 51 being washed and prepared for transfer to Tank 40. The acceptance evaluation needs to be completed prior to the transfer of the material in Tank 51 to Tank 40 to complete the formation of SB5. The sludge slurry in Tank 40 has already been qualified for DWPF and is currently being processed as SB4. The radionuclide concentrations were measured or estimated in the Tank 51 SB5 Qualification Sample prepared at Savannah River National Laboratory (SRNL). This sample was prepared from the three liter sample of Tank 51 sludge slurry taken on March 21, 2008. The sample was delivered to SRNL where it was initially characterized in the Shielded Cells. Under direction of the Liquid Waste Organization it was then modified by five washes, six decants, an addition of Pu/Be from Canyon Tank 16.4, and an addition of NaNO2. This final slurry now has a composition expected to be similar to that of the slurry in Tank 51 after final preparations have been made for transfer of that slurry to Ta Determining the radionuclide concentrations in this Tank 51 SB5 Qualification Sample is part of the work requested in Technical Task Request (TTR) No. HLW-DWPF-TTR-2008-0010. The work with this qualification sample is covered by a Task Technical and Quality Assurance Plan and an Analytical Study Plan. The radionuclides included in this report are needed for the DWPF Radiological Program Evaluation, the DWPF Waste Acceptance Criteria (TSR/WAC) Evaluation, and the DWPF Solid Waste Characterization Program (TTR Task 2). Radionuclides required to meet the Waste Acceptance Product Specifications (TTR Task 5) will be measured at a later date after the slurry from Tank 51 has been transferred to Tank 40. Then a sample of the as-processed SB5 will be taken and

Evaluation of FTIR-based analytical methods for the analysis of simulated wastes

International Nuclear Information System (INIS)

Rebagay, T.V.; Cash, R.J.; Dodd, D.A.; Lockrem, L.L.; Meacham, J.E.; Winkelman, W.D.

1994-01-01

Three FTIR-based analytical methods that have potential to characterize simulated waste tank materials have been evaluated. These include: (1) fiber optics, (2) modular transfer optic using light guides equipped with non-contact sampling peripherals, and (3) photoacoustic spectroscopy. Pertinent instrumentation and experimental procedures for each method are described. The results show that the near-infrared (NIR) region of the infrared spectrum is the region of choice for the measurement of moisture in waste simulants. Differentiation of the NIR spectrum, as a preprocessing steps, will improve the analytical result. Preliminary data indicate that prominent combination bands of water and the first overtone band of the ferrocyanide stretching vibration may be utilized to measure water and ferrocyanide species simultaneously. Both near-infrared and mid-infrared spectra must be collected, however, to measure ferrocyanide species unambiguously and accurately. For ease of sample handling and the potential for field or waste tank deployment, the FTIR-Fiber Optic method is preferred over the other two methods. Modular transfer optic using light guides and photoacoustic spectroscopy may be used as backup systems and for the validation of the fiber optic data

High-level core sample x-ray imaging at the Hanford Site

International Nuclear Information System (INIS)

Weber, J.R.; Keve, J.K.

1995-10-01

Waste tank sampling of radioactive high-level waste is required for continued operations, waste characterization, and site safety. Hanford Site tank farms consist of 28 double-shell and 149 single-shell underground storage tanks. The single shell tanks are out-of-service an no longer receive liquid waste. Core samples of salt cake and sludge waste are remotely obtained using truck-mounted, core drill platforms. Samples are recovered from tanks through a 2.25 inch (in.) drill pipe in 26-in. steel tubes, 1.5 in. diameter. Drilling parameters vary with different waste types. Because sample recovery has been marginal an inadequate at times, a system was needed to provide drill truck operators with ''real-time feedback'' about the physical condition of the sample and the percent recovery, prior to making nuclear assay measurements and characterizations at the analytical laboratory. The Westinghouse Hanford Company conducted proof-of-principal radiographic testing to verify the feasibility of a proposed imaging system. Tests were conducted using an iridium 192 radiography source to determine the effects of high radiation on image quality. The tests concluded that samplers with a dose rate in excess of 5000 R/hr could be imaged with only a slight loss of image quality and samples less than 1000 R/hr have virtually no effect on image quality. The Mobile Core Sample X-Ray Examination System, a portable vendor-engineered assembly, has components uniquely configured to produce a real-time radiographic system suitable for safely examining radioactive tank core segments collected at the Hanford Site. The radiographic region of interest extends from the bottom (valve) of the sampler upward 19 to 20 in. The purpose of the Mobile Core Sample X-Ray Examination System is to examine the physical contents of core samples after removal from the tank and prior to placement in an onsite transfer cask

Heat transfer analysis of the waste-container sleeve/salt configuration

International Nuclear Information System (INIS)

Callahan, G.D.; Ratigan, J.L.; Russell, J.E.; Fossum, A.F.

1975-01-01

Prior to this investigation, the heat transport considered was only that of straight conduction. The waste container, air gap, and sleeve arrangement was considered to be a single, consistent, time-dependent, heat-generating unit in intimate contact with the salt. The conduction model does not accurately model the heat transfer mechanisms available. Thus radiation and combined radiation and convection must also be considered in the determination of the temperature field. As would be expected, the canister temperatures are higher for the case of radiation across the airgap than those that result from conduction when the canister is in intimate contact with the salt. For the radiation case, the canister temperatures rise rapidly to a temperature of approximately 1,140 0 F and maintain an almost steady state condition for one year whereafter the temperatures slowly decrease. The far field temperatures, near the pillar centerline, are essentially equivalent for all cases. As time proceeds, the far field temperatures of the conduction models are about 15% different

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

Radioactive waste storage facility and underground disposal method for radioactive wastes using the facility

International Nuclear Information System (INIS)

Endo, Yoshihiro.

1997-01-01

A sealed container storage chamber is formed in underground rocks. A container storage pool is formed on the inner bottom of the sealed vessel storage chamber. A heat exchanger for cooling water and a recycling pump are disposed on an operation floor of the sealed vessel storage chamber. Radioactive wastes sealed vessels in which radioactive wastes are sealed are transferred from the ground to the sealed vessel storage chamber through a sealed vessel transferring shaft, and immersed in cooling water stored in the vessel storage pool. When after heat of the radioactive wastes is removed by the cooling water, the cooling water in the vessel storage pool is sucked up to the ground surface. After dismantling equipments, bentonite-type fillers are filled in the inside of the sealed vessel storage chamber, sealed vessel transferring shaft, air supplying shaft and air exhaustion shaft, and the radioactive waste-sealed vessels can be subjected stably to into underground disposal. (I.N.)

Parametric optimization and heat transfer analysis of a dual loop ORC (organic Rankine cycle) system for CNG engine waste heat recovery

International Nuclear Information System (INIS)

Yang, Fubin; Zhang, Hongguang; Yu, Zhibin; Wang, Enhua; Meng, Fanxiao; Liu, Hongda; Wang, Jingfu

2017-01-01

In this study, a dual loop ORC (organic Rankine cycle) system is adopted to recover exhaust energy, waste heat from the coolant system, and intercooler heat rejection of a six-cylinder CNG (compressed natural gas) engine. The thermodynamic, heat transfer, and optimization models for the dual loop ORC system are established. On the basis of the waste heat characteristics of the CNG engine over the whole operating range, a GA (genetic algorithm) is used to solve the Pareto solution for the thermodynamic and heat transfer performances to maximize net power output and minimize heat transfer area. Combined with optimization results, the optimal parameter regions of the dual loop ORC system are determined under various operating conditions. Then, the variation in the heat transfer area with the operating conditions of the CNG engine is analyzed. The results show that the optimal evaporation pressure and superheat degree of the HT (high temperature) cycle are mainly influenced by the operating conditions of the CNG engine. The optimal evaporation pressure and superheat degree of the HT cycle over the whole operating range are within 2.5–2.9 MPa and 0.43–12.35 K, respectively. The optimal condensation temperature of the HT cycle, evaporation and condensation temperatures of the LT (low temperature) cycle, and exhaust temperature at the outlet of evaporator 1 are kept nearly constant under various operating conditions of the CNG engine. The thermal efficiency of the dual loop ORC system is within the range of 8.79%–10.17%. The dual loop ORC system achieves the maximum net power output of 23.62 kW under the engine rated condition. In addition, the operating conditions of the CNG engine and the operating parameters of the dual loop ORC system significantly influence the heat transfer areas for each heat exchanger. - Highlights: • A dual loop ORC system is adopted to recover the waste heat of a CNG engine. • Parametric optimization and heat transfer analysis are

Waste Tank Organic Safety Program: Analytical methods development. Progress report, FY 1994

International Nuclear Information System (INIS)

Campbell, J.A.; Clauss, S.A.; Grant, K.E.

1994-09-01

The objectives of this task are to develop and document extraction and analysis methods for organics in waste tanks, and to extend these methods to the analysis of actual core samples to support the Waste Tank organic Safety Program. This report documents progress at Pacific Northwest Laboratory (a) during FY 1994 on methods development, the analysis of waste from Tank 241-C-103 (Tank C-103) and T-111, and the transfer of documented, developed analytical methods to personnel in the Analytical Chemistry Laboratory (ACL) and 222-S laboratory. This report is intended as an annual report, not a completed work

Arc plasma incineration of surrogate radioactive wastes

International Nuclear Information System (INIS)

Girold, C.; Cartier, R.; Taupiac, J.P.; Vandensteendam, C.; Baronnet, J.M.

1995-01-01

The aim of this presentation is to demonstrate the feasibility to substitute a single plasma reactor, where the arc is transferred on a melt glass bath, for several steps in an existing nuclear technological wastes incinerator. The incineration of wastes, the produced gas treatment and the vitrification of ashes issued from waste incineration are the three simultaneous functions of this new kind of reactor. The three steps of the work are described: first, post-combustion in an oxygen plasma of gases generated from the waste pyrolysis, then, vitrification of ashes from the calcination of wastes in the transferred plasma furnace and finally, incineration/vitrification of wastes in the same furnace

Environmental sampling program for a solar evaporation pond for liquid radioactive wastes

International Nuclear Information System (INIS)

Romero, R.; Gunderson, T.C.; Talley, A.D.

1980-04-01

Los Alamos Scientific Laboratory (LASL) is evaluating solar evaporation as a method for disposal of liquid radioactive wastes. This report describes a sampling program designed to monitor possible escape of radioactivity to the environment from a solar evaporation pond prototype constructed at LASL. Background radioactivity levels at the pond site were determined from soil and vegetation analyses before construction. When the pond is operative, the sampling program will qualitatively and quantitatively detect the transport of radioactivity to the soil, air, and vegetation in the vicinity. Possible correlation of meteorological data with sampling results is being investigated and measures to control export of radioactivity by biological vectors are being assessed

Bioethanol production from waste bread samples made from mixtures of wheat and buckwheat flours

OpenAIRE

Ačanski, Marijana; Pastor, Kristian; Razmovski, Radojka; Vučurović, Vesna; Psodorov, Đorđe

2014-01-01

In this paper yields of bioethanol from seven samples of bread were compared. Samples of bread were produced and prepared in a laboratory by mixing wheat and buckwheat flour in amounts of 0, 20, 40, 50, 60, 80 and 100%. At first, the analysis of all seven samples of bread was done (dry matter, starch content and pH value of bread sample suspensions). Then the waste bread suspensions were hydrolyzed by applying commercial hydrolytic enzymes, Termamyl® SC and SAN Extra® L. The fermentation proc...

Preparation and analysis of standardized waste samples for Controlled Ecological Life Support Systems (CELSS)

Science.gov (United States)

Carden, J. L.; Browner, R.

1982-01-01

The preparation and analysis of standardized waste samples for controlled ecological life support systems (CELSS) are considered. Analysis of samples from wet oxidation experiments, the development of ion chromatographic techniques utilizing conventional high pressure liquid chromatography (HPLC) equipment, and an investigation of techniques for interfacing an ion chromatograph (IC) with an inductively coupled plasma optical emission spectrometer (ICPOES) are discussed.

The effects of aging on compressive strength of low-level radioactive waste form samples

International Nuclear Information System (INIS)

McConnell, J.W. Jr.; Neilson, R.M. Jr.

1996-06-01

The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program, funded by the US Nuclear Regulatory Commission (NRC), is (a) studying the degradation effects in organic ion-exchange resins caused by radiation, (b) examining the adequacy of test procedures recommended in the Branch Technical Position on Waste Form to meet the requirements of 10 CFR 61 using solidified ion-exchange resins, (c) obtaining performance information on solidified ion-exchange resins in a disposal environment, and (d) determining the condition of liners used to dispose ion-exchange resins. Compressive tests were performed periodically over a 12-year period as part of the Technical Position testing. Results of that compressive testing are presented and discussed. During the study, both portland type I-II cement and Dow vinyl ester-styrene waste form samples were tested. This testing was designed to examine the effects of aging caused by self-irradiation on the compressive strength of the waste forms. Also presented is a brief summary of the results of waste form characterization, which has been conducted in 1986, using tests recommended in the Technical Position on Waste Form. The aging test results are compared to the results of those earlier tests. 14 refs., 52 figs., 5 tabs

Hanford Waste Vitrification Plant full-scale feed preparation testing with water and process simulant slurries

International Nuclear Information System (INIS)

Gaskill, J.R.; Larson, D.E.; Abrigo, G.P.

1996-03-01

The Hanford Waste Vitrification Plant was intended to convert selected, pretreated defense high-level waste and transuranic waste from the Hanford Site into a borosilicate glass. A full-scale testing program was conducted with nonradioactive waste simulants to develop information for process and equipment design of the feed-preparation system. The equipment systems tested included the Slurry Receipt and Adjustment Tank, Slurry Mix Evaporator, and Melter-Feed Tank. The areas of data generation included heat transfer (boiling, heating, and cooling), slurry mixing, slurry pumping and transport, slurry sampling, and process chemistry. 13 refs., 129 figs., 68 tabs

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.

Characterization of high level nuclear waste glass samples following extended melter idling

Energy Technology Data Exchange (ETDEWEB)

Fox, Kevin M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Peeler, David K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kruger, Albert A. [USDOE Office of River Protection, Richland, WA (United States)

2015-06-16

The Savannah River Site Defense Waste Processing Facility (DWPF) melter was recently idled with glass remaining in the melt pool and riser for approximately three months. This situation presented a unique opportunity to collect and analyze glass samples since outages of this duration are uncommon. The objective of this study was to obtain insight into the potential for crystal formation in the glass resulting from an extended idling period. The results will be used to support development of a crystal-tolerant approach for operation of the high-level waste melter at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Two glass pour stream samples were collected from DWPF when the melter was restarted after idling for three months. The samples did not contain crystallization that was detectible by X-ray diffraction. Electron microscopy identified occasional spinel and noble metal crystals of no practical significance. Occasional platinum particles were observed by microscopy as an artifact of the sample collection method. Reduction/oxidation measurements showed that the pour stream glasses were fully oxidized, which was expected after the extended idling period. Chemical analysis of the pour stream glasses revealed slight differences in the concentrations of some oxides relative to analyses of the melter feed composition prior to the idling period. While these differences may be within the analytical error of the laboratories, the trends indicate that there may have been some amount of volatility associated with some of the glass components, and that there may have been interaction of the glass with the refractory components of the melter. These changes in composition, although small, can be attributed to the idling of the melter for an extended period. The changes in glass composition resulted in a 70-100 °C increase in the predicted spinel liquidus temperature (TL) for the pour stream glass samples relative to the analysis of the melter feed prior to

METHODS FOR DETERMINING AGITATOR MIXING REQUIREMENTS FOR A MIXING & SAMPLING FACILITY TO FEED WTP (WASTE TREATMENT PLANT)

Energy Technology Data Exchange (ETDEWEB)

GRIFFIN PW

2009-08-27

The following report is a summary of work conducted to evaluate the ability of existing correlative techniques and alternative methods to accurately estimate impeller speed and power requirements for mechanical mixers proposed for use in a mixing and sampling facility (MSF). The proposed facility would accept high level waste sludges from Hanford double-shell tanks and feed uniformly mixed high level waste to the Waste Treatment Plant. Numerous methods are evaluated and discussed, and resulting recommendations provided.

Transfer of radioactive waste management expertise from Switzerland to other countries with small nuclear power programmes

International Nuclear Information System (INIS)

McKinley, I.; Birkhaeuser, Ph.; Kickmaier, W.; Vomvoris, S.; Zuidema, P.

2000-01-01

A legal requirement which coupled demonstration of the feasibility of nuclear waste disposal to the extension of reactor operational licenses beyond 1985 acted to force rapid development of the Swiss radioactive waste management programme. Over a period of almost 30 years and at a cost of approximately 800 M CHF Nagra has become established as a centre of excellence in this field. Resources include highly experienced manpower, literature and databases supporting development of national repositories for L/ILW and HLW/TRU and state-of-the-art R and D infrastructure (including 2 underground laboratories, hot-laboratory facilities at PSI (Paul Scherrer Institute), modelling groups at universities etc.). This paper reviews Nagra's experience and considers various ways in which expertise can be transferred to other small countries to minimise duplication of effort and optimise development of their own national programmes. (author)

Transfer function design based on user selected samples for intuitive multivariate volume exploration

KAUST Repository

Zhou, Liang

2013-02-01

Multivariate volumetric datasets are important to both science and medicine. We propose a transfer function (TF) design approach based on user selected samples in the spatial domain to make multivariate volumetric data visualization more accessible for domain users. Specifically, the user starts the visualization by probing features of interest on slices and the data values are instantly queried by user selection. The queried sample values are then used to automatically and robustly generate high dimensional transfer functions (HDTFs) via kernel density estimation (KDE). Alternatively, 2D Gaussian TFs can be automatically generated in the dimensionality reduced space using these samples. With the extracted features rendered in the volume rendering view, the user can further refine these features using segmentation brushes. Interactivity is achieved in our system and different views are tightly linked. Use cases show that our system has been successfully applied for simulation and complicated seismic data sets. © 2013 IEEE.

Transfer function design based on user selected samples for intuitive multivariate volume exploration

KAUST Repository

Zhou, Liang; Hansen, Charles

2013-01-01

Multivariate volumetric datasets are important to both science and medicine. We propose a transfer function (TF) design approach based on user selected samples in the spatial domain to make multivariate volumetric data visualization more accessible for domain users. Specifically, the user starts the visualization by probing features of interest on slices and the data values are instantly queried by user selection. The queried sample values are then used to automatically and robustly generate high dimensional transfer functions (HDTFs) via kernel density estimation (KDE). Alternatively, 2D Gaussian TFs can be automatically generated in the dimensionality reduced space using these samples. With the extracted features rendered in the volume rendering view, the user can further refine these features using segmentation brushes. Interactivity is achieved in our system and different views are tightly linked. Use cases show that our system has been successfully applied for simulation and complicated seismic data sets. © 2013 IEEE.

Long-term interactions of full-scale cemented waste simulates with salt brines

Energy Technology Data Exchange (ETDEWEB)

Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

2016-07-01

Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO{sub 3} solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

Long-term interactions of full-scale cemented waste simulates with salt brines

International Nuclear Information System (INIS)

Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

2016-01-01

Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO 3 solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

Transfer Stations for urban solid wastes. Design; Transferencia ferroviaria de residuos solidos domiciliarios. Alternativas para el diseno

Energy Technology Data Exchange (ETDEWEB)

Dussuel Jurado, E.

2004-07-01

On the basis of experience in the design of projects promoted by Spanish companies for Transfer Stations for Solid Domestic Waste (RSD) using the railway, aspects regarding location, technological alternatives, restrictions and dimensions are considered. The equipment necessary for the different options is described. The paper is concerned with the Metropolitan Area of Santiago (Chile) and comments on and discusses the regulatory restrictions laid down by the PRMS (Metropolitan Regulatory Plan for Santiago) affecting decisions on location, dimensions and design. Last of all, the costs of Transfer Stations are analysed depending on the distance from the sanitary landfill nd in comparison with the alternative by road, laying down convenience kerbs. Also analysed are the environmental advantages of the railway option as opposed to roads. (Author)

Aerospace vehicle water-waste management

Science.gov (United States)

Pecoraro, J. N.

1973-01-01

The collection and disposal of human wastes, such as urine and feces, in a spacecraft environment are performed in an aesthetic and reliable manner to prevent degradation of crew performance. The waste management system controls, transfers, and processes materials such as feces, emesis, food residues, used expendables, and other wastes. The requirements, collection, transport, and waste processing are described.

Sequential determination of important ecotoxic radionuclides in nuclear waste samples

International Nuclear Information System (INIS)

Bilohuscin, J.

2016-01-01

In the dissertation thesis we focused on the development and optimization of a sequential determination method for radionuclides 93 Zr, 94 Nb, 99 Tc and 126 Sn, employing extraction chromatography sorbents TEVA (R) Resin and Anion Exchange Resin, supplied by Eichrom Industries. Prior to the attestation of sequential separation of these proposed radionuclides from radioactive waste samples, a unique sequential procedure of 90 Sr, 239 Pu, 241 Am separation from urine matrices was tried, using molecular recognition sorbents of AnaLig (R) series and extraction chromatography sorbent DGA (R) Resin. On these experiments, four various sorbents were continually used for separation, including PreFilter Resin sorbent, which removes interfering organic materials present in raw urine. After the acquisition of positive results of this sequential procedure followed experiments with a 126 Sn separation using TEVA (R) Resin and Anion Exchange Resin sorbents. Radiochemical recoveries obtained from samples of radioactive evaporate concentrates and sludge showed high efficiency of the separation, while values of 126 Sn were under the minimum detectable activities MDA. Activity of 126 Sn was determined after ingrowth of daughter nuclide 126m Sb on HPGe gamma detector, with minimal contamination of gamma interfering radionuclides with decontamination factors (D f ) higher then 1400 for 60 Co and 47000 for 137 Cs. Based on the acquired experiments and results of these separation procedures, a complex method of sequential separation of 93 Zr, 94 Nb, 99 Tc and 126 Sn was proposed, which included optimization steps similar to those used in previous parts of the dissertation work. Application of the sequential separation method for sorbents TEVA (R) Resin and Anion Exchange Resin on real samples of radioactive wastes provided satisfactory results and an economical, time sparing, efficient method. (author)

Characterization of Samples from the Effluent Treatment Facility Evaporator Waste Concentrate Tank

Energy Technology Data Exchange (ETDEWEB)

Wilmarth, W.R. [Westinghouse Savannah River Company, AIKEN, SC (United States)

1998-01-31

During October 1997, the ETF Evaporator Waste Concentrate Tank No. 2 was discovered to contain a significant accumulation of solid deposits. SRTC performed destructive and nondestructive examination of solid samples from the tank. The results of these tests indicate that the solids contain mixtures of sodium oxalate (65 percent), the sulfide enclathrated sodium aluminosilicate (30 percent), and iron oxide (5 percent).

Characterization of Samples from the Effluent Treatment Facility Evaporator Waste Concentrate Tank

International Nuclear Information System (INIS)

Wilmarth, W.R.

1998-01-01

During October 1997, the ETF Evaporator Waste Concentrate Tank No. 2 was discovered to contain a significant accumulation of solid deposits. SRTC performed destructive and nondestructive examination of solid samples from the tank. The results of these tests indicate that the solids contain mixtures of sodium oxalate (65 percent), the sulfide enclathrated sodium aluminosilicate (30 percent), and iron oxide (5 percent)

Radioactive waste material melter apparatus

Science.gov (United States)

Newman, D.F.; Ross, W.A.

1990-04-24

An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another. 8 figs.

Radioactive waste material melter apparatus

International Nuclear Information System (INIS)

Newman, D.F.; Ross, W.A.

1990-01-01

An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another. 8 figs

Effect of heterogeneity and anisotropy related to the construction method on transfer processes in waste rock piles.

Science.gov (United States)

Lahmira, Belkacem; Lefebvre, René; Aubertin, Michel; Bussière, Bruno

2016-01-01

Waste rock piles producing acid mine drainage (AMD) are partially saturated systems involving multiphase (gas and liquid) flow and coupled transfer processes. Their internal structure and heterogeneous properties are inherited from their wide-ranging material grain sizes, their modes of deposition, and the underlying topography. This paper aims at assessing the effect of physical heterogeneity and anisotropy of waste rock piles on the physical processes involved in the generation of AMD. Generic waste rock pile conditions were represented with the numerical simulator TOUGH AMD based on those found at the Doyon mine waste rock pile (Canada). Models included four randomly distributed material types (coarse, intermediate, fine and very fine-grained). The term "randomly" as used in this study means that the vertical profile and spatial distribution of materials in waste rock piles (internal structure) defy stratigraphy principles applicable to natural sediments (superposition and continuity). The materials have different permeability and capillary properties, covering the typical range of materials found in waste rock piles. Anisotropy with a larger horizontal than vertical permeability was used to represent the effect of pile construction by benches, while the construction by end-dumping was presumed to induce a higher vertical than horizontal permeability. Results show that infiltrated precipitation preferentially flows in fine-grained materials, which remain almost saturated, whereas gas flows preferentially through the most permeable coarse materials, which have higher volumetric gas saturation. Anisotropy, which depends on pile construction methods, often controls global gas flow paths. Construction by benches favours lateral air entry close to the pile slope, whereas end-dumping leads to air entry from the surface to the interior of the pile by secondary gas convection cells. These results can be useful to construct and rehabilitate waste rock piles to minimize

Mass transfer from penetrations in waste containers

International Nuclear Information System (INIS)

Pescatore, C.; Sastre, C.

1987-01-01

Recent studies have indicated that localized corrosion of a relatively small area of a waste container may impair the containment function to such an extent that larger releases may be possible than from the bare waste form. This would take place when a large number of holes coexist on the container while their concentration fields do not interact significantly with each other. After performing a steady state analysis of the release from a hole, it is shown that much fewer independent holes can coexist on a container surface than previously estimated. The calculated radionuclide release from multiple independent holes must be changed accordingly. Previous analyses did not proceed to a correct application of the linear superposition principle. This resulted in unacceptable physical conclusions and undue strain on the performance assessment necessary for a container licensing procedure. The paper also analyzes the steady state release from penetrations of finite length and whose concentration fields interact with one another. The predicted release from these penetrations is lower than the previously calculated release from holes of zero thickness. It is concluded here that the steady-state release from multiple holes on a waste container can not exceed the release from the bare waste form and that multiple perforations need not be a serious liability to container performance. 8 refs., 3 figs., 1 tab

Analysis of the effects of explosion of a hydrogen cylinder on the transfer of radioactive liquid wastes at nuclear power stations

International Nuclear Information System (INIS)

Lopes, Karina B.; Melo, Paulo Fernando F.F. e

2011-01-01

This work presents a study of explosion effects of a stored hydrogen cylinder on the transfer of radioactive liquid wastes at nuclear power plants. The peak overpressure is calculated, as well as the strength of resulting fragments, thus confirming the main harmful effect of an explosion of flammable vapor cloud, based on the TNT equivalent method. The scenarios identified are calculated and compared with the overpressure ranges of 1%, 50% and 99% of structural damages, which were determined by the Eisenberg's vulnerability model. The results show that the overpressure and the resulting fragments from the explosion of a hydrogen gas cylinder are not able to cause the overturning of the tanker under study, and also show that a minimum distance of 30 meters between the hydrogen cylinder and the tanker can be considered a safe distance to the passage of this tanker during the transfer of radioactive liquid waste, in which the likelihood of occurrence of structural damages is less than 1%. (author)

Double Shell Tank (DST) Transfer Pump Subsystem Specification

International Nuclear Information System (INIS)

LESHIKAR, G.A.

2000-01-01

This specification establishes the performance requirements and provides references to the requisite codes and standards to be applied to the Double-Shell Tank (DST) Transfer Pump Subsystem which supports the first phase of Waste Feed Delivery (WFD). This specification establishes the performance requirements and provides the references to the requisite codes and standards to be applied during the design of the DST Transfer Pump Subsystem that supports the first phase of (WFD). The DST Transfer Pump Subsystem consists of a pump for supernatant and or slurry transfer for the DSTs that will be retrieved during the Phase 1 WFD operations. This system is used to transfer low-activity waste (LAW) and high-level waste (HLW) to designated DST staging tanks. It also will deliver blended LAW and HLW feed from these staging tanks to the River Protection Project (RPP) Privatization Contractor facility where it will be processed into an immobilized waste form. This specification is intended to be the basis for new projects/installations (W-521, etc.). This specification is not intended to retroactively affect previously established project design criteria without specific direction by the program

Fiscal Year 2001 Tank Characterization Technical Sampling Basis and Waste Information Requirements Document

International Nuclear Information System (INIS)

ADAMS, M.R.

2000-01-01

The Fiscal Year 2001 Tank Characterization Technical Sampling Basis and Waste Information Requirements Document (TSB-WIRD) has the following purposes: (1) To identify and integrate sampling and analysis needs for fiscal year (FY) 2001 and beyond. (2) To describe the overall drivers that require characterization information and to document their source. (3) To describe the process for identifying, prioritizing, and weighting issues that require characterization information to resolve. (4) To define the method for determining sampling priorities and to present the sampling priorities on a tank-by-tank basis. (5) To define how the characterization program is going to satisfy the drivers, close issues, and report progress. (6)To describe deliverables and acceptance criteria for characterization deliverables

Prediction of centrifugal pump-cleaning ability in waste sludge

International Nuclear Information System (INIS)

Churnetski, B.V.

1981-01-01

Radioactive waste at the Savannah River Plant (SRP) is being transferred from older waste tanks to new, stress-relieved tanks for more effective waste management. The technology developed for waste removal involves the use of long-shaft, recirculating, centrifugal pumps (slurry pumps). Testing completed at the Savannah River Laboratory's 30-meter-diameter mock-up waste tank related the effective cleaning radius (ECR) of a slurry pump to critical pump and materials characteristics. Presently, this theory is being applied to radioactive waste at SRP. However, the technology can be applied to other remote handling situations where the slurry rheology can be determined. For SRP waste, an equation of the form: ECR α DV 0 (rho/tau 0 )/sup 1/2/ was determined where D is the nozzle diameter, V 0 is the average initial velocity, rho is the density of the slurry, and tau 0 is the yield stress of the slurry. Using this relationship, the cleaning performance of a pump operating in any SRP sludge environment can be predicted. Specifically, yield stress and density measurements on sludge samples can be used to predict the required number and effective location for slurry pumps in actual SRP waste tanks

Application to transfer radioactive waste to the Nevada Test Site

International Nuclear Information System (INIS)

1992-01-01

All waste described in this application has been, and will be, generated by LANL in support of the nuclear weapons test program at the NTS. All waste originates on the NTS. DOE Order 5820.2A states that low-level radioactive waste shall be disposed of at the site where it is generated, when practical. Since the waste is produced at the NTS, it is cost effective for LANL to dispose of the waste at the NTS

Grout treatment facility dangerous waste permit application

International Nuclear Information System (INIS)

1992-07-01

The Grout Treatment Facility (GTF) will provide permanent disposal for approximately 43 Mgal of radioactive liquid waste currently being stored in underground tanks on the Hanford Site. The first step in permanent disposal is accomplished by solidifying the low-level liquid waste with cementitious dry materials. The resulting grout is cast within underground vaults. This report on the GTF contains information on the following: Hanford Site Maps, road evaluation for the grout treatment facility, Department of Ecology certificate of non-designation for centralia fly ash, double-shell tank waste compositional modeling, laboratory analysis reports for double-shell tank waste, stored in tanks 241-AN-103, 241-AN-106, and 241-AW-101, grout vault heat transfer results for M-106 grout formulation, test results for extraction procedure toxicity testing, test results for toxicity testing of double-shell tank grout, pilot-scale grout production test with a simulated low-level waste, characterization of simulated low-level waste grout produced in a pilot-scale test, description of the procedure for sampling nonaging waste storage tanks, description of laboratory procedures, grout campaign waste composition verification, variability in properties of grouted phosphate/sulfate N-reactor waste, engineering drawings, description of operating procedures, equipment list--transportable grout equipment, grout treatment facility--tank integrity assessment plan, long-term effects of waste solutions on concrete and reinforcing steel, vendor information, grout disposal facilities construction quality assurance plan, and flexible membrane liner/waste compatibility test results

Thermophysical and heat transfer properties of phase change material candidate for waste heat transportation system

Science.gov (United States)

Kaizawa, Akihide; Maruoka, Nobuhiro; Kawai, Atsushi; Kamano, Hiroomi; Jozuka, Tetsuji; Senda, Takeshi; Akiyama, Tomohiro

2008-05-01

A waste heat transportation system trans-heat (TH) system is quite attractive that uses the latent heat of a phase change material (PCM). The purpose of this paper is to study the thermophysical properties of various sugars and sodium acetate trihydrate (SAT) as PCMs for a practical TH system and the heat transfer property between PCM selected and heat transfer oil, by using differential scanning calorimetry (DSC), thermogravimetry-differential thermal analysis (TG-DTA) and a heat storage tube. As a result, erythritol, with a large latent heat of 344 kJ/kg at melting point of 117°C, high decomposition point of 160°C and excellent chemical stability under repeated phase change cycles was found to be the best PCM among them for the practical TH system. In the heat release experiments between liquid erythritol and flowing cold oil, we observed foaming phenomena of encapsulated oil, in which oil droplet was coated by solidification of PCM.

Tank vapor sampling and analysis data package for tank 241-C-106 waste retrieval sluicing system process test phase III

Energy Technology Data Exchange (ETDEWEB)

LOCKREM, L.L.

1999-08-13

This data package presents sampling data and analytical results from the March 28, 1999, vapor sampling of Hanford Site single-shell tank 241-C-106 during active sluicing. Samples were obtained from the 296-C-006 ventilation system stack and ambient air at several locations. Characterization Project Operations (CPO) was responsible for the collection of all SUMMATM canister samples. The Special Analytical Support (SAS) vapor team was responsible for the collection of all triple sorbent trap (TST), sorbent tube train (STT), polyurethane foam (PUF), and particulate filter samples collected at the 296-C-006 stack. The SAS vapor team used the non-electrical vapor sampling (NEVS) system to collect samples of the air, gases, and vapors from the 296-C-006 stack. The SAS vapor team collected and analyzed these samples for Lockheed Martin Hanford Corporation (LMHC) and Tank Waste Remediation System (TWRS) in accordance with the sampling and analytical requirements specified in the Waste Retrieval Sluicing System Vapor Sampling and Analysis Plan (SAP) for Evaluation of Organic Emissions, Process Test Phase III, HNF-4212, Rev. 0-A, (LMHC, 1999). All samples were stored in a secured Radioactive Materials Area (RMA) until the samples were radiologically released and received by SAS for analysis. The Waste Sampling and Characterization Facility (WSCF) performed the radiological analyses. The samples were received on April 5, 1999.

METHODS FOR DETERMINING AGITATOR MIXING REQUIREMENTS FOR A MIXING and SAMPLING FACILITY TO FEED WTP (WASTE TREATMENT PLANT)

International Nuclear Information System (INIS)

Griffin, P.W.

2009-01-01

The following report is a summary of work conducted to evaluate the ability of existing correlative techniques and alternative methods to accurately estimate impeller speed and power requirements for mechanical mixers proposed for use in a mixing and sampling facility (MSF). The proposed facility would accept high level waste sludges from Hanford double-shell tanks and feed uniformly mixed high level waste to the Waste Treatment Plant. Numerous methods are evaluated and discussed, and resulting recommendations provided.

A hazard and probabilistic safety analysis of a high-level waste transfer process

International Nuclear Information System (INIS)

Bott, T.F.; Sasser, M.K.

1996-01-01

This paper describes a safety analysis of a transfer process for high-level radioactive and toxic waste. The analysis began with a hazard assessment that used elements of What If, Checklist, Failure Modes and Effects Analysis, and Hazards and Operability Study (HAZOP) techniques to identify and rough-in accident sequences. Based on this preliminary analysis, the most significant accident sequences were developed further using event trees. Quantitative frequency estimates for the accident sequences were based on operational data taken from the historical record of the site where the process is performed. Several modeling challenges were encountered in the course of the study. These included linked initiating and accident progression events, fire propagation modeling, accounting for administrative control violations, and handling mission-phase effects

History of waste tank 22, 1965--1974

International Nuclear Information System (INIS)

McNatt, F.G.

1979-04-01

Tank 22 (a 1,300,000-gallon Type IV tank) was placed in service June 6, 1965, receiving HW from tank 21. The HW was transferred back into tank 21 in September 1965 and fed to the Building 242-H evaporator. This recycled concentrate and concentrate from other waste was then received in tank 22 until the tank was filled. The HW concentrate and salt remained in the tank until November 1971 when removal was begun. The concentrated supernate was transferred from the tank followed by dissolution and removal of salt from the tank walls and bottom. The salt removal was completed in May 1974 and since that time tank 22 has served as a receiver of LW from Building 221-H. Inspections of the tank interior were made using a 40-ft optical periscope and the steel thickness of the tank bottom was measured ultrasonically. Samples of the tank vapors and liquid collected in the sidewall and bottom sumps were analyzed. Temperature and specific gravity measurements were made of waste stored in the tank. Several equipment modifications and repairs were made

Waste Preparation and Transport Chemistry: Results of the FY 2002 Studies

Energy Technology Data Exchange (ETDEWEB)

Hunt, R.D.

2003-07-10

The initial step in the remediation of nuclear waste stored at Hanford and the Savannah River Site (SRS) involves the retrieval and transfer of the waste to another tank or to a treatment facility. The retrieved waste can range from a filtered supernatant to a slurry. Nearly all of the recent solid formation problems encountered during waste transfers and subsequent treatment steps have involved decanted or filtered supernatants. Problems with slurry transfers have not yet surfaced, because tank farm operations at Hanford and the SRS have focused primarily on supernatant transfers and treatment. For example, the interim stabilization program at Hanford continues to reduce the level of supernatants and interstitial liquids in its single-shell tanks through saltwell pumping of filtered liquid. In addition, at present, the cross-site transfer lines at Hanford can be used only to transfer liquids. Another reason for fewer problems with slurry transfers involves the additions of large quantities of dilution water prior to the transfer. When the waste is transferred, a drop in temperature is expected because most transfer lines are not heated. However, the dilution water reduces or eliminates solid formation caused by this temperature drop. In sharp contrast, decanted or filtered supernatants are near or at saturation for certain compounds. In such cases, tank farm operators must continue to evaporate their liquid waste since available tank space is quite limited. Solid formation can occur when the temperature of saturated solutions drops even slightly. The evaporation step can also lead to the formation of problematic solids. At the SRS, the evaporation of a relatively dilute waste stream was suspended due to the formation of deposits in the evaporator system. Therefore, small drops in temperature or evaporation can lead to problematic solid formations.

Waste Preparation and Transport Chemistry: Results of the FY 2002 Studies

International Nuclear Information System (INIS)

Hunt, R.D.

2003-01-01

The initial step in the remediation of nuclear waste stored at Hanford and the Savannah River Site (SRS) involves the retrieval and transfer of the waste to another tank or to a treatment facility. The retrieved waste can range from a filtered supernatant to a slurry. Nearly all of the recent solid formation problems encountered during waste transfers and subsequent treatment steps have involved decanted or filtered supernatants. Problems with slurry transfers have not yet surfaced, because tank farm operations at Hanford and the SRS have focused primarily on supernatant transfers and treatment. For example, the interim stabilization program at Hanford continues to reduce the level of supernatants and interstitial liquids in its single-shell tanks through saltwell pumping of filtered liquid. In addition, at present, the cross-site transfer lines at Hanford can be used only to transfer liquids. Another reason for fewer problems with slurry transfers involves the additions of large quantities of dilution water prior to the transfer. When the waste is transferred, a drop in temperature is expected because most transfer lines are not heated. However, the dilution water reduces or eliminates solid formation caused by this temperature drop. In sharp contrast, decanted or filtered supernatants are near or at saturation for certain compounds. In such cases, tank farm operators must continue to evaporate their liquid waste since available tank space is quite limited. Solid formation can occur when the temperature of saturated solutions drops even slightly. The evaporation step can also lead to the formation of problematic solids. At the SRS, the evaporation of a relatively dilute waste stream was suspended due to the formation of deposits in the evaporator system. Therefore, small drops in temperature or evaporation can lead to problematic solid formations

Low-impact sampling under an active solid low-level radioactive waste disposal unit using horizontal drilling technology

International Nuclear Information System (INIS)

Puglisi, C.V.; Vold, E.L.

1995-01-01

The purpose of this project was to determine the performance of the solid low-level radioactive waste (LLRW) disposal units located on a mesa top at TA-54, Area G, Los Alamos National Laboratory (LANL), Los Alamos, NM, and to provide in-situ (vadose zone) site characterization information to Area G's Performance Assessment. The vadose zone beneath an active disposal unit (DU 37), was accessed by utilizing low-impact, air-rotary horizontal drilling technology. Core samples were pulled, via wire-line core method, in 3 horizontal holes fanning out below DU 37 at approximately 5 foot intervals depending on recovery percentage. Samples were surveyed and prepared in-field following Environmental Restoration (ER) guidelines. Samples were transferred from the field to the CST-9 Radvan for initial radiological screening. Following screening, samples were delivered to CST-3 analytical lab for analyses including moisture content, 23 inorganics, 60 volatile organic compounds (VOC's), 68 semivolatile organic compounds (SVOC's), tritium, lead 210, radium 226 ampersand 228, cesium 137, isotopic plutonium, americium 241, strontium 90, isotopic uranium, and isotopic thorium. Other analyses included matric potential, alpha spectroscopy, gamma spectroscopy, and gross alpha/beta. The overall results of the analysis identified only tritium as having migrated from the DU. Am-241, Eu-152, and Pu-238 were possibly identified above background but the results are not definitive. Of all organics analysed for, only ethyl acetate was tentatively identified slightly above background. All inorganics were found to be well below regulatory limits. Based on the results of the above mentioned analyses, it was determined that Area G's disposal units are performing well and no significant liquid phase migration of contaminants has occurred

Contents and Sample Arguments of a Safety Case for Near Surface Disposal of Radioactive Waste

International Nuclear Information System (INIS)

2017-06-01

This publication arises from the results of two projects to assist Member States in understanding and developing safety cases for near-surface radioactive waste disposal facilities. The objective of the publication is to give detailed information on the contents of safety cases for radioactive waste disposal and the types of arguments that may be included. It is written for technical experts preparing a safety case, and decision makers in the regulatory body and government. The publication outlines the key uses and aspects of the safety case, its evolution in parallel with that of the disposal facility, the key decision steps in the development of the waste disposal facility, the components of the safety case, their place in the Matrix of Arguments for a Safety Case (the MASC matrix), and a detailed description of the development of sample arguments that might be included in a safety case for each of two hypothetical radioactive waste disposal facilities.

Exergo-economic analysis of finned tube for waste heat recovery including phase change heat transfer

International Nuclear Information System (INIS)

Wu, Shuang Ying; Jiu, Jing Rui; Xiao, Lan; Li, You Rong; Liu, Chao; Xu, Jin Liang

2013-01-01

In this paper, an exergo-economic criterion, i.e. the net profit per unit transferred heat load, is established from the perspective of exergy recovery to evaluate the performance of finned tube used in waste heat recovery. Also, the dimensionless exergy change number is introduced to investigate the effect of the flow (mechanical) exergy loss rate on the recovered thermal exergy. Selecting R245fa as a working fluid and exhaust flue gas as a heat source, the effects of the internal Reynolds number Re_i, the external Reynolds number Re_o , the unit cost of thermal exergy ε_q , the geometric parameter of finned tube η_oβ and the phase change temperature T_v etc. on the performance of finned tube are discussed in detail. The results show that the higher T_v and η_oβ, and lower Re_i may lead to the negligible flow(mechanical) exergy loss rate. There exists an optimal value of Re_i where the net profit per unit transferred heat load peaks, while the variations of Re_o, ε_q and T_v cause monotonic change of the net profit per unit transferred heat load. The phase change temperature exerts relatively greater influence on the exergo-economic performance of finned tube in comparison with other parameters. And there exists a critical phase change temperature, where the net profit per unit transferred heat load is equal to zero.

Nuclear waste storage container with metal matrix

Science.gov (United States)

Sump, Kenneth R.

1978-01-01

The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties.

Nuclear waste storage container with metal matrix

International Nuclear Information System (INIS)

Sump, K.R.

1978-01-01

The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties

Impact of cementitious materials decalcification on transfer properties: application to radioactive waste deep repository; Influence de la decalcification de materiaux cimentaires sur les proprietes de transfert: application au stockage profond de dechets radioactifs

Energy Technology Data Exchange (ETDEWEB)

Perlot, C

2005-09-15

Cementitious materials have been selected to compose the engineering barrier system (EBS) of the French radioactive waste deep repository, because of concrete physico-chemical properties: the hydrates of the cementitious matrix and the pH of the pore solution contribute to radionuclides retention; furthermore the compactness of these materials limits elements transport. The confinement capacity of the system has to be assessed while a period at least equivalent to waste activity (up to 100.000 years). His durability was sustained by the evolution of transfer properties in accordance with cementitious materials decalcification, alteration that expresses structure long-term behavior. Then, two degradation modes were carried out, taking into account the different physical and chemical solicitations imposed by the host formation. The first mode, a static one, was an accelerated decalcification test using nitrate ammonium solution. It replicates the EBS alteration dues to underground water. Degradation kinetic was estimated by the amount of calcium leached and the measurement of the calcium hydroxide dissolution front. To evaluate the decalcification impact, samples were characterized before and after degradation in term of microstructure (porosity, pores size distribution) and of transfer properties (diffusivity, gas and water permeability). The influence of cement nature (ordinary Portland cement, blended cement) and aggregates type (lime or siliceous) was observed: experiments were repeated on different mortars mixes. On this occasion, an essential reflection on this test metrology was led. The second mode, a dynamical degradation, was performed with an environmental permeameter. It recreates the EBS solicitations ensured during the re-saturation period, distinguished by the hydraulic pressure imposed by the geologic layer and the waste exothermicity. This apparatus, based on triaxial cell functioning, allows applying on samples pressure drop between 2 and 10 MPa and

Radioactive Waste Repositories Administration - SURAO

International Nuclear Information System (INIS)

Kucerka, M.

1998-01-01

The Atomic Act specifies, among other things, responsibilities of the government in the field of safe disposal of radioactive wastes. To satisfy this responsibility, the Ministry of Industry and Trade has established the Radioactive Waste Repositories Administration (SURAO). SURAO's major responsibilities include: (a) the preparation, construction, commissioning, operation, and decommissioning of radioactive waste repositories and the monitoring of their environmental impacts; (b) radioactive waste management; (c) spent or irradiated nuclear fuel processing into a form suitable for storage/disposal or reuse; (d) record-keeping of received radioactive wastes and their producers; (e) administration of fund transfers as stipulated by the Atomic Act, Article 27; (f) development of proposals for specification of fees to be paid to the Nuclear Account; (g) responsibility for and coordination of research and development in the field of radioactive waste handling and management; (h) supervision of licensees' margin earmarked for the decommissioning of their facilities; (i) providing services in radioactive waste handling and management; (j) handling and management of radioactive wastes that have been transferred to the Czech Republic from abroad and cannot be sent back; (k) interim administration of radioactive wastes that have become state property. The Statute of the Administration is reproduced in full. (P.A.)

WTP Waste Feed Qualification: Glass Fabrication Unit Operation Testing Report

Energy Technology Data Exchange (ETDEWEB)

Stone, M. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Hanford Missions Programs; Newell, J. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Process Technology Programs; Johnson, F. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Engineering Process Development; Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Engineering Process Development

2016-07-14

The waste feed qualification program is being developed to protect the Hanford Tank Waste Treatment and Immobilization Plant (WTP) design, safety basis, and technical basis by assuring waste acceptance requirements are met for each staged waste feed campaign prior to transfer from the Tank Operations Contractor to the feed receipt vessels inside the Pretreatment Facility. The Waste Feed Qualification Program Plan describes the three components of waste feed qualification: 1. Demonstrate compliance with the waste acceptance criteria 2. Determine waste processability 3. Test unit operations at laboratory scale. The glass fabrication unit operation is the final step in the process demonstration portion of the waste feed qualification process. This unit operation generally consists of combining each of the waste feed streams (high-level waste (HLW) and low-activity waste (LAW)) with Glass Forming Chemicals (GFCs), fabricating glass coupons, performing chemical composition analysis before and after glass fabrication, measuring hydrogen generation rate either before or after glass former addition, measuring rheological properties before and after glass former addition, and visual observation of the resulting glass coupons. Critical aspects of this unit operation are mixing and sampling of the waste and melter feeds to ensure representative samples are obtained as well as ensuring the fabrication process for the glass coupon is adequate. Testing was performed using a range of simulants (LAW and HLW simulants), and these simulants were mixed with high and low bounding amounts of GFCs to evaluate the mixing, sampling, and glass preparation steps in shielded cells using laboratory techniques. The tests were performed with off-the-shelf equipment at the Savannah River National Laboratory (SRNL) that is similar to equipment used in the SRNL work during qualification of waste feed for the Defense Waste Processing Facility (DWPF) and other waste treatment facilities at the

Characterization, Leaching, and Filtration Testing for Bismuth Phosphate Sludge (Group 1) and Bismuth Phosphate Saltcake (Group 2) Actual Waste Sample Composites

International Nuclear Information System (INIS)

Lumetta, Gregg J.; Buck, Edgar C.; Daniel, Richard C.; Draper, Kathryn; Edwards, Matthew K.; Fiskum, Sandra K.; Hallen, Richard T.; Jagoda, Lynette K.; Jenson, Evan D.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Shimskey, Rick W.; Sinkov, Sergey I.; Snow, Lanee A.

2009-01-01

A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan.() The test program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. The actual waste-testing program included homogenizing the samples by group, characterizing the solids and aqueous phases, and performing parametric leaching tests. Two of the eight defined groups - bismuth phosphate sludge (Group 1) and bismuth phosphate saltcake (Group 2) - are the subjects of this report. The Group 1 waste was anticipated to be high in phosphorus and was implicitly assumed to be present as BiPO4 (however, results presented here indicate that the phosphate in Group 1 is actually present as amorphous iron(III) phosphate). The Group 2 waste was also anticipated to be high in phosphorus, but because of the relatively low bismuth content and higher aluminum content, it was anticipated that the Group 2 waste would contain a mixture of gibbsite, sodium phosphate, and aluminum phosphate. Thus, the focus of the Group 1 testing was on determining the behavior of P removal during caustic leaching, and the focus of the Group 2 testing was on the removal of both P and Al. The waste-type definition, archived sample conditions, homogenization activities, characterization (physical, chemical, radioisotope, and crystal habit), and caustic leaching behavior as functions of time, temperature, and hydroxide concentration are discussed in this report. Testing was conducted according to TP-RPP-WTP-467

Taiwan industrial cooperation program technology transfer for low-level radioactive waste final disposal - phase I.

Energy Technology Data Exchange (ETDEWEB)

Knowlton, Robert G.; Cochran, John Russell; Arnold, Bill Walter; Jow, Hong-Nian; Mattie, Patrick D.; Schelling, Frank Joseph Jr. (; .)

2007-01-01

Sandia National Laboratories and the Institute of Nuclear Energy Research, Taiwan have collaborated in a technology transfer program related to low-level radioactive waste (LLW) disposal in Taiwan. Phase I of this program included regulatory analysis of LLW final disposal, development of LLW disposal performance assessment capabilities, and preliminary performance assessments of two potential disposal sites. Performance objectives were based on regulations in Taiwan and comparisons to those in the United States. Probabilistic performance assessment models were constructed based on limited site data using software including GoldSim, BLT-MS, FEHM, and HELP. These software codes provided the probabilistic framework, container degradation, waste-form leaching, groundwater flow, radionuclide transport, and cover infiltration simulation capabilities in the performance assessment. Preliminary performance assessment analyses were conducted for a near-surface disposal system and a mined cavern disposal system at two representative sites in Taiwan. Results of example calculations indicate peak simulated concentrations to a receptor within a few hundred years of LLW disposal, primarily from highly soluble, non-sorbing radionuclides.

Tank waste remediation system dangerous waste training plan

International Nuclear Information System (INIS)

POHTO, R.E.

1999-01-01

This document outlines the dangerous waste training program developed and implemented for all Treatment, Storage, and Disposal (TSD) Units operated by Lockheed Martin Hanford Corporation (LMHC) Tank Waste Remediation System (TWRS) in the Hanford 200 East, 200 West and 600 Areas and the <90 Day Accumulation Area at 209E. Operating TSD Units operated by TWRS are: the Double-Shell Tank (DST) System (including 204-AR Waste Transfer Building), the 600 Area Purgewater Storage and the Effluent Treatment Facility. TSD Units undergoing closure are: the Single-Shell Tank (SST) System, 207-A South Retention Basin, and the 216-B-63 Trench

Mixing Modeling Analysis For SRS Salt Waste Disposition

International Nuclear Information System (INIS)

Lee, S.

2011-01-01

Nuclear waste at Savannah River Site (SRS) waste tanks consists of three different types of waste forms. They are the lighter salt solutions referred to as supernate, the precipitated salts as salt cake, and heavier fine solids as sludge. The sludge is settled on the tank floor. About half of the residual waste radioactivity is contained in the sludge, which is only about 8 percentage of the total waste volume. Mixing study to be evaluated here for the Salt Disposition Integration (SDI) project focuses on supernate preparations in waste tanks prior to transfer to the Salt Waste Processing Facility (SWPF) feed tank. The methods to mix and blend the contents of the SRS blend tanks were evalutaed to ensure that the contents are properly blended before they are transferred from the blend tank such as Tank 50H to the SWPF feed tank. The work consists of two principal objectives to investigate two different pumps. One objective is to identify a suitable pumping arrangement that will adequately blend/mix two miscible liquids to obtain a uniform composition in the tank with a minimum level of sludge solid particulate in suspension. The other is to estimate the elevation in the tank at which the transfer pump inlet should be located where the solid concentration of the entrained fluid remains below the acceptance criterion (0.09 wt% or 1200 mg/liter) during transfer operation to the SWPF. Tank 50H is a Waste Tank that will be used to prepare batches of salt feed for SWPF. The salt feed must be a homogeneous solution satisfying the acceptance criterion of the solids entrainment during transfer operation. The work described here consists of two modeling areas. They are the mixing modeling analysis during miscible liquid blending operation, and the flow pattern analysis during transfer operation of the blended liquid. The modeling results will provide quantitative design and operation information during the mixing/blending process and the transfer operation of the blended

WASTES: a waste management logistics/economics model

International Nuclear Information System (INIS)

McNair, G.W.; Shay, M.R.; Fletcher, J.F.; Cashwell, J.W.

1985-02-01

The WASTES model simulates a user defined system for nuclear waste transportation and storage at both temporary and long-term storage facilities. The model is written in FORTRAN 77 as an extension to the SLAM commercial simulation package (Pritsker and Pegden 1979). SLAM (Simulation Language for Alternative Modeling) is utilized in a discrete event mode to model the passage of spent fuel through the system. The system is initiated with individual reactor discharges of spent fuel as described in the reactor discharge data file or as supplied by the user. The reactor discharge file contains deterministic information on the date (year/month) and quantity of spent fuel discharges. From this point, the model is controlled by a combination of source originated and destination originated transfers. Source driven transfers occur when a reactor pool violates the full core reserve (FCR) storage margin or when the reactor is decommissioned. At these times, the source reactor checks destination facilities to see if they can accept material. A dry storage facility is assumed to exist for each reactor and is allowed to grow as necessary to contain spent fuel which cannot be shipped to any other facility. In this way the FCR margin is always maintained. Destination driven transfers occur when the annual capacity of a facility will not be met by full core reserve or decommissioning shipments. An attempt is made at the end of each calendar year to schedule enough shipments of spent fuel from facilities with non-critical storage capacity to fill the annual capacity of each destination facility. Allowable facility types are reprocessing plants, federal interim storage (FIS), monitored retrievable storage (MRS), and repositories. The number, capacities, location and priority for receipt of spent fuel is user specified. This report describes in detail the waste generating model, the waste facilities model, the transportation model and the basic transportation scheme

Enhanced Waste Tank Level Model

Energy Technology Data Exchange (ETDEWEB)

Duignan, M.R.