User Manual for the AZ-101 Data Acquisition System (AS-101 DAS)

Energy Technology Data Exchange (ETDEWEB)

BRAYTON, D.D.

2000-02-17

User manual for the TK AZ-101 Waste Retrieval System Data Acquisition System. The purpose of this document is to describe use of the AZ-101 Data Acquisition System (AZ-101 DAS). The AZ-101 DAS is provided to fulfill the requirements for data collection and monitoring as defined in Letters of Instruction (LOI) from Numatec Hanford Corporation (NHC) to Fluor Federal Services (FFS). For a complete description of the system, including design, please refer to the AZ-101 DAS System Description document, RPP-5572.

User Manual for the AZ-101 Data Acquisition System (AS-101 DAS)

International Nuclear Information System (INIS)

BRAYTON, D.D.

2000-01-01

User manual for the TK AZ-101 Waste Retrieval System Data Acquisition System. The purpose of this document is to describe use of the AZ-101 Data Acquisition System (AZ-101 DAS). The AZ-101 DAS is provided to fulfill the requirements for data collection and monitoring as defined in Letters of Instruction (LOI) from Numatec Hanford Corporation (NHC) to Fluor Federal Services (FFS). For a complete description of the system, including design, please refer to the AZ-101 DAS System Description document, RPP-5572

Comparison of simulants to actual neutralized current acid waste: process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

Energy Technology Data Exchange (ETDEWEB)

Morrey, E.V.; Tingey, J.M.; Elliott, M.L.

1996-10-01

A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs were established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste was performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property ,models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

Comparison of simulants to actual neutralized current acid waste: Process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

Energy Technology Data Exchange (ETDEWEB)

Morrey, E.V.; Tingey, J.M.

1996-04-01

A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs have been established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste is being performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

Comparison of simulants to actual neutralized current acid waste: process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

International Nuclear Information System (INIS)

Morrey, E.V.; Tingey, J.M.; Elliott, M.L.

1996-10-01

A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs were established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste was performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property ,models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions

Comparison of simulants to actual neutralized current acid waste: Process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

International Nuclear Information System (INIS)

Morrey, E.V.; Tingey, J.M.

1996-04-01

A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs have been established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste is being performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions

Dose rate analysis for Tank 101 AZ (Project W151)

International Nuclear Information System (INIS)

Schwarz, R.A.; Hillesland, K.E.; Carter, L.L.

1994-11-01

This document describes the expected dose rates for modification to tank 101 AZ including modifications to the steam coil, mixer pump, and temperature probes. The thrust of the effort is to determine dose rates from: modification of a steam coil and caisson; the installation of mixer pumps; the installation of temperature probes; and estimates of dose rates that will be encountered while making these changes. Because the dose rates for all of these configurations depend upon the photon source within the supernate and sludge, comparisons were also made between measured dose rates within a drywell and the corresponding calculated dose rates. The calculational tool used is a Monte Carlo (MCNP 2 ) code since complicated three dimensional geometries are involved. A summary of the most important results of the entire study is given in Section 2. The basic calculational geometry model of the tank is discussed in Section 3, along with a tabulation of the photon sources that were used within the supernate and the sludge, and a discussion of uncertainties. The calculated dose rates around the steam coil and caisson before and after modification are discussed in Section 4. The configuration for the installation of the mixer pumps and the resulting dose rates are given in Section 5. The predicted changes in dose rates due to a possible dilution of the supernate source are given in Section 6. The calculational configuration used to model the installation of temperature probes and the resulting predicted dose rates are discussed in Section 7. Finally, comparisons of measured to calculated dose rates within a drywell are summarized in Section 8. Extended discussions of calculational models and Monte Carlo optimization techniques used are included in Appendix A

Evaluation of AY/AZ tank farm ventilation system during aging waste retrieval operations

International Nuclear Information System (INIS)

Wong, J.J.; Waters, E.D.

1995-01-01

Waste Management is currently planning to demonstrate mobilization of radioactive waste sludges in Tank 101-AZ beginning in October 1991. The retrieval system being designed will utilize mixer pumps that generate high-velocity, high-volume submerged liquid jets to mobilize settled solids. There is concern that these jets may also generate radioactive aerosols, some of which may be carried into the tank Ventilation system. The purpose of this study is to determine if the current AY/AZ ventilation system or the proposed ventilation system upgrade (Project W-030) will provide adequate deentrainment of liquid and solid aerosols during mixer pump operations, or if the radioactive aerosols will overload the HEPA filters

Evaluation of waste temperatures in AWF tanks for bypass mode operation of the 702-AZ ventilation system (Project W-030)

International Nuclear Information System (INIS)

Sathyanarayana, K.

1997-01-01

This report describes the results of thermal hydraulic analysis performed to provide data in support of Project W-030 to startup new 702-AZ Primary Ventilation System. During the startup of W-030 system, the ventilation system will be operating in bypass mode. In bypass made of operation, the system is capable of supplying 1000 cfm total flow for all four AWF doubleshell tanks. The design of the W-030 system is based on the assumption that both the recirculation loop of the primary ventilation system and the secondary ventilation which provides cooling would be operating. However, during the startup neither the recirculation system nor the secondary ventilation system will be operating. A minimum flow of 100 cfm is required to prevent any flammable gas associated risk. The remaining 600 cfm flow can be divided among the four tanks as necessary to keep the peak sludge temperatures below the operating temperature limit. For the purpose of determining the minimum flow required for cooling each tank, the thermal hydraulic analysis is performed to predict the peak sludge temperatures in AY/AZ tanks under different ventilation flows. The heat load for AZ farm tanks is taken from characterization reports and for the AY farm tanks, the heat load was estimated by thermal analysis using the measured waste temperatures and the waste liquid evaporation rates. The tank 241-AZ-101 and the tank 241-AZ-102 have heat loads of 241,600 and 199,500 Btu/hr respectively. The tank 241-AY-101 and tank 241-AY-102 have heat loads of 41,000 and 33,000 Btu/hr respectively. Using the ambient meteorological conditions of temperature and relative humidity for the air and tank, some soil surface and the sludge levels reported in recent documents, the peak sludge and supernatant temperatures were predicted for various primary ventilation flows ranging from 100 to 400 cfm for AZ tanks and 100 and 150 cfm for AY tanks. The results of these thermal hydraulic analyses are presented. Based on the

241-AY/AZ waste storage tanks: Supplemental gravity load analysis. Volume 1

International Nuclear Information System (INIS)

Baliga, R.

1994-01-01

An analysis of the 241SY tanks performed by ADVENT(1994b) to resolve dome overload issues indicated that the tank can sustain the dome loads resulting from additional soil overburden depth, increased soil density, and increased concentrated load. Similar issues exist for the 241AY/AZ tanks and therefore, an interim analysis of the 241AY/AZ tanks is presented herein. The scope of this effort is to review and compare all design drawings pertaining to the 241AY and 241AZ tanks with those pertaining to the 241SY tanks; to modify the axisymmetric model of the 241SY tanks to represent the 241AY/AZ tanks; and to evaluate the effect of additional dome load on the 241AY/AZ tanks by performing a structural analysis for gravity loads (dead load + live load). ADVENTS's additional scope of work is to perform a qualitative evaluation of the 241AY/AZ tanks for seismic and thermal loadings (Vollert 1982 and Blume 1971). This qualitative evaluation does not include any detailed finite element analysis of the tanks. The following design-based gravity loading conditions are used in this interim analysis of the 241AY/AZ tanks to determine a baseline for the tank stresses or section loads

AZ-101 Mixer Pump Demonstration and Tests Data Management Analysis Plan

Energy Technology Data Exchange (ETDEWEB)

DOUGLAS, D.G.

2000-02-22

This document provides a plan for the analysis of the data collected during the AZ-101 Mixer Pump Demonstration and Tests. This document was prepared after a review of the AZ-101 Mixer Pump Test Plan (Revision 4) [1] and other materials. The plan emphasizes a structured and well-ordered approach towards handling and examining the data. This plan presumes that the data will be collected and organized into a unified body of data, well annotated and bearing the date and time of each record. The analysis of this data will follow a methodical series of steps that are focused on well-defined objectives. Section 2 of this plan describes how the data analysis will proceed from the real-time monitoring of some of the key sensor data to the final analysis of the three-dimensional distribution of suspended solids. This section also identifies the various sensors or sensor systems and associates them with the various functions they serve during the test program. Section 3 provides an overview of the objectives of the AZ-101 test program and describes the data that will be analyzed to support that test. The objectives are: (1) to demonstrate that the mixer pumps can be operated within the operating requirements; (2) to demonstrate that the mixer pumps can mobilize the sludge in sufficient quantities to provide feed to the private contractor facility, and (3) to determine if the in-tank instrumentation is sufficient to monitor sludge mobilization and mixer pump operation. Section 3 also describes the interim analysis that organizes the data during the test, so the analysis can be more readily accomplished. Section 4 describes the spatial orientation of the various sensors in the tank. This section is useful in visualizing the relationship of the Sensors in terms of their location in the tank and how the data from these sensors may be related to the data from other sensors. Section 5 provides a summary of the various analyses that will be performed on the data during the test

AZ-101 Mixer Pump Demonstration and Tests: Data Management (Analysis) Plan

International Nuclear Information System (INIS)

DOUGLAS, D.G.

2000-01-01

This document provides a plan for the analysis of the data collected during the AZ-101 Mixer Pump Demonstration and Tests. This document was prepared after a review of the AZ-101 Mixer Pump Test Plan (Revision 4) [1] and other materials. The plan emphasizes a structured and well-ordered approach towards handling and examining the data. This plan presumes that the data will be collected and organized into a unified body of data, well annotated and bearing the date and time of each record. The analysis of this data will follow a methodical series of steps that are focused on well-defined objectives. Section 2 of this plan describes how the data analysis will proceed from the real-time monitoring of some of the key sensor data to the final analysis of the three-dimensional distribution of suspended solids. This section also identifies the various sensors or sensor systems and associates them with the various functions they serve during the test program. Section 3 provides an overview of the objectives of the AZ-101 test program and describes the data that will be analyzed to support that test. The objectives are: (1) to demonstrate that the mixer pumps can be operated within the operating requirements; (2) to demonstrate that the mixer pumps can mobilize the sludge in sufficient quantities to provide feed to the private contractor facility, and (3) to determine if the in-tank instrumentation is sufficient to monitor sludge mobilization and mixer pump operation. Section 3 also describes the interim analysis that organizes the data during the test, so the analysis can be more readily accomplished. Section 4 describes the spatial orientation of the various sensors in the tank. This section is useful in visualizing the relationship of the Sensors in terms of their location in the tank and how the data from these sensors may be related to the data from other sensors. Section 5 provides a summary of the various analyses that will be performed on the data during the test

Results of Sludge Mobilization Testing at Hanford High Level Waste (HLW) Tank

International Nuclear Information System (INIS)

STAEHR, T.W.

2001-01-01

Waste stored in the Tank 241-AZ-101 at the US DOE Hanford is scheduled as the initial feed for high-level waste vitrification. Tank 241-AZ-101 currently holds over 3,000,000 liters of waste made up of a settled sludge layer covered by a layer of liquid supernant. To retrieve the waste from the tank, it is necessary to mobilize and suspend the settled sludge so that the resulting slurry can be pumped from the tank for treatment and vitrification. Two 223.8-kilowatt mixer pumps have been installed in Tank 241-AZ-101 to mobilize the settled sludge layer of waste for retrieval. In May of 2000, the mixer pumps were subjected to a series of tests to determine (1) the extent to which the mixer pumps could mobilize the settle sludge layer of waste, (2) if the mixer pumps could function within operating parameters, and (3) if state-of-the-art monitoring equipment could effectively monitor and quantify the degree of sludge mobilization and suspension. This paper presents the major findings and results of the Tank 241-AZ-101 mixer pump tests, based on analysis of data and waste samples that were collected during the testing. Discussion of the results focuses on the effective cleaning radius achieved and the volume and concentration of sludge mobilized, with both one and two pumps operating in various configurations and speeds. The Tank 241-AZ-101 mixer pump tests were unique in that sludge mobilization parameters were measured using actual waste in an underground storage tank at the hanford Site. The methods and instruments that were used to measure waste mobilization parameters in Tank 241-AZ-101 can be used in other tanks. It can be concluded from the testing that the use of mixer pumps is an effective retrieval method for the mobilization of settled solids in Tank 241-AZ-101

AZ-101 Mixer Pump Test Qualification Test Procedures (QTP)

International Nuclear Information System (INIS)

THOMAS, W.K.

2000-01-01

Describes the Qualification test procedure for the AZ-101 Mixer Pump Data Acquisition System (DAS). The purpose of this Qualification Test Procedure (QTP) is to confirm that the AZ-101 Mixer Pump System has been properly programmed and hardware configured correctly. This QTP will test the software setpoints for the alarms and also check the wiring configuration from the SIMcart to the HMI. An Acceptance Test Procedure (ATP), similar to this QTP will be performed to test field devices and connections from the field

Tank 241-AW-101 tank characterization plan

International Nuclear Information System (INIS)

Sathyanarayana, P.

1994-01-01

The first section gives a summary of the available information for Tank AW-101. Included in the discussion are the process history and recent sampling events for the tank, as well as general information about the tank such as its age and the risers to be used for sampling. Tank 241-AW-101 is one of the 25 tanks on the Flammable Gas Watch List. To resolve the Flammable Gas safety issue, characterization of the tanks, including intrusive tank sampling, must be performed. Prior to sampling, however, the potential for the following scenarios must be evaluated: the potential for ignition of flammable gases such as hydrogen-air and/or hydrogen-nitrous oxide; and the potential for secondary ignition of organic-nitrate/nitrate mixtures in crust layer initiated by the burning of flammable gases or by a mechanical in-tank energy source. The characterization effort applicable to this Tank Characterization Plan is focused on the resolution of the crust burn flammable gas safety issue of Tank AW-101. To evaluate the potential for a crust burn of the waste material, calorimetry tests will be performed on the waste. Differential Scanning Calorimetry (DSC) will be used to determine whether an exothermic reaction exists

241-AY-101 Tank Construction Extent of Condition Review for Tank Integrity

Energy Technology Data Exchange (ETDEWEB)

Barnes, Travis J.; Gunter, Jason R.

2013-08-26

This report provides the results of an extent of condition construction history review for tank 241-AY-101. The construction history of tank 241-AY-101 has been reviewed to identify issues similar to those experienced during tank AY-102 construction. Those issues and others impacting integrity are discussed based on information found in available construction records, using tank AY-102 as the comparison benchmark. In tank 241-AY-101, the second double-shell tank constructed, similar issues as those with tank 241-AY-102 construction reoccurred. The overall extent of similary and affect on tank 241-AY-101 integrity is described herein.

Operational Test Report for the 241-AZ-101 Suspended Solids Profiler

International Nuclear Information System (INIS)

STENKAMP, D.M.

2000-01-01

This document comprises the Operational Test Report for the 241-AZ-101 Suspended Solids Profiler. This document presents the results of Operational Testing of the 241-AZ-101 Suspended Solids Profiler (SSP). Testing of the SSP was performed in accordance with OTP-260-005, ''SUSPENDED SOLIDS PROFILER OPERATIONAL TEST PROCEDURE''. The objective of the testing was to verify that all equipment and components functioned as designed, following construction completion and turnover to operations

Tank 241-C-101 vapor sampling and analysis tank characterization report

International Nuclear Information System (INIS)

Huckaby, J.L.

1995-01-01

Tank C-101 headspace gas and vapor samples were collected and analyzed to help determine the potential risks of fugitive emissions to tank farm workers. Gas and vapor samples from the Tank C-101 headspace were collected on July 7, 1994 using the in situ sampling (ISS) method, and again on September 1, 1994 using the more robust vapor sampling system (VSS). Gas and vapor concentrations in Tank C-101 are influenced by its connections to other tanks and its ventilation pathways. At issue is whether the organic vapors in Tank C-101 are from the waste in that tank, or from Tanks C-102 or C-103. Tank C-103 is on the Organic Watch List; the other two are not. Air from the Tank C-101 headspace was withdrawn via a 7.9-m long heated sampling probe mounted in riser 8, and transferred via heated tubing to the VSS sampling manifold. The tank headspace temperature was determined to be 34.0 C, and all heated zones of the VSS were maintained at approximately 50 C. Sampling media were prepared and analyzed by WHC, Oak Ridge National Laboratories, Pacific Northwest Laboratories, and Oregon Graduate Institute of Science and Technology through a contract with Sandia National Laboratories. The 39 tank air samples and 2 ambient air control samples collected are listed in Table X-1 by analytical laboratory. Table X-1 also lists the 14 trip blanks and 2 field blanks provided by the laboratories

Acceptance Test Report for the 241-AZ-101 Ultrasonic Interface Level Analyzer

International Nuclear Information System (INIS)

ANDREWS, J.E.

2000-01-01

This document comprises the Acceptance Test Report for the 241-AZ-101 Ultrasonic Interface Level Analyzer. This document presents the results of Acceptance Testing of the 241-AZ-101 Ultrasonic Interface Level Analyzers (URSILLAs). Testing of the URSILLAs was performed in accordance with ATP-260-001, ''URSILLA Pre-installation Acceptance Test Procedure''. The objective of the testing was to verify that all equipment and components function in accordance with design specifications and original equipment manufacturer's specifications

Project W-211 Initial Tank Retrieval Systems (ITRS) Description of Operations for 241-AZ-102

Energy Technology Data Exchange (ETDEWEB)

BRIGGS, S.R.

2000-02-25

The primary purpose of the Initial Tank Retrieval Systems (ITRS) is to provide systems for retrieval of radioactive wastes stored in underground double-shell tanks (DSTs) for transfer to alternate storage, evaporation, pretreatment or treatment, while concurrently reducing risks associated with safety watch list and other DSTs. This Description of Operation (DOO) defines the control philosophy for the waste retrieval system for Tank 241-AZ-102 (AZ-102). This DOO provides a basis for the detailed design of the Project W-211 Retrieval Control System (RCS) for AZ-102 and also establishes test criteria for the RCS.

Project W-211 Initial Tank Retrieval Systems (ITRS) Description of Operations for 241-AZ-102

International Nuclear Information System (INIS)

BRIGGS, S.R.

2000-01-01

The primary purpose of the Initial Tank Retrieval Systems (ITRS) is to provide systems for retrieval of radioactive wastes stored in underground double-shell tanks (DSTs) for transfer to alternate storage, evaporation, pretreatment or treatment, while concurrently reducing risks associated with safety watch list and other DSTs. This Description of Operation (DOO) defines the control philosophy for the waste retrieval system for Tank 241-AZ-102 (AZ-102). This DOO provides a basis for the detailed design of the Project W-211 Retrieval Control System (RCS) for AZ-102 and also establishes test criteria for the RCS

49 CFR 179.101 - Individual specification requirements applicable to pressure tank car tanks.

Science.gov (United States)

2010-10-01

... to pressure tank car tanks. 179.101 Section 179.101 Transportation Other Regulations Relating to... MATERIALS REGULATIONS SPECIFICATIONS FOR TANK CARS Specifications for Pressure Tank Car Tanks (Classes DOT... tank car tanks. Editorial Note: At 66 FR 45186, Aug. 28, 2001, an amendment published amending a table...

Tank 241-AZ-102 Privatization Push Mode Core Sampling and Analysis Plan

International Nuclear Information System (INIS)

RASMUSSEN, J.H.

1999-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AZ-102. The purpose of this sampling event is to obtain information about the characteristics of the contents of 241-AZ-102 required to satisfy the Data Quality Objectives For TWRS Privatization Phase I : Confirm Tank TIS An Appropriate Feed Source For High-Level Waste Feed Batch X(HLW DQO) (Nguyen 1999a), Data Quality Objectives For TWRS Privatization Phase 1: Confirm Tank TIS An Appropriate Feed Source For Low-Activity Waste Feed Batch X (LAW DQO) (Nguyen 1999b), Low Activity Waste and High Level Waste Feed Data Quality Objectives (L and H DQO) (Patello et al. 1999) and Characterization Data Needs for Development, Design, and Operation of Retrieval Equipment Developed through the Data Quality Objective Process (Equipment DQO) (Bloom 1996). The Tank Characterization Technical Sampling Basis document (Brown et al. 1998) indicates that these issues, except the Equipment DQO apply to tank 241-AZ-102 for this sampling event. The Equipment DQO is applied for shear strength measurements of the solids segments only. Poppiti (1999) requires additional americium-241 analyses of the sludge segments. Brown et al. (1998) also identify safety screening, regulatory issues and provision of samples to the Privatization Contractor(s) as applicable issues for this tank. However, these issues will not be addressed via this sampling event. Reynolds et al. (1999) concluded that information from previous sampling events was sufficient to satisfy the safety screening requirements for tank 241-AZ-102. Push mode core samples will be obtained from risers 15C and 24A to provide sufficient material for the chemical analyses and tests required to satisfy these data quality objectives. The 222-S Laboratory will extrude core samples, composite the liquids and solids, perform chemical analyses

Acceptance Test Report for the 241-AZ-101 Ultrasonic Interface Level Analyzer

Energy Technology Data Exchange (ETDEWEB)

ANDREWS, J.E.

2000-01-27

This document comprises the Acceptance Test Report for the 241-AZ-101 Ultrasonic Interface Level Analyzer. This document presents the results of Acceptance Testing of the 241-AZ-101 Ultrasonic Interface Level Analyzers (URSILLAs). Testing of the URSILLAs was performed in accordance with ATP-260-001, ''URSILLA Pre-installation Acceptance Test Procedure''. The objective of the testing was to verify that all equipment and components function in accordance with design specifications and original equipment manufacturer's specifications.

Double-Shell Tank Visual Inspection Changes Resulting from the Tank 241-AY-102 Primary Tank Leak

International Nuclear Information System (INIS)

Girardot, Crystal L.; Washenfelder, Dennis J.; Johnson, Jeremy M.; Engeman, Jason K.

2013-01-01

As part of the Double-Shell Tank (DST) Integrity Program, remote visual inspections are utilized to perform qualitative in-service inspections of the DSTs in order to provide a general overview of the condition of the tanks. During routine visual inspections of tank 241-AY-102 (AY-102) in August 2012, anomalies were identified on the annulus floor which resulted in further evaluations. In October 2012, Washington River Protection Solutions, LLC determined that the primary tank of AY-102 was leaking. Following identification of the tank AY-102 probable leak cause, evaluations considered the adequacy of the existing annulus inspection frequency with respect to the circumstances of the tank AY-102 1eak and the advancing age of the DST structures. The evaluations concluded that the interval between annulus inspections should be shortened for all DSTs, and each annulus inspection should cover > 95 percent of annulus floor area, and the portion of the primary tank (i.e., dome, sidewall, lower knuckle, and insulating refractory) that is visible from the annulus inspection risers. In March 2013, enhanced visual inspections were performed for the six oldest tanks: 241-AY-101, 241-AZ-101,241-AZ-102, 241-SY-101, 241-SY-102, and 241-SY-103, and no evidence of leakage from the primary tank were observed. Prior to October 2012, the approach for conducting visual examinations of DSTs was to perform a video examination of each tank's interior and annulus regions approximately every five years (not to exceed seven years between inspections). Also, the annulus inspection only covered about 42 percent of the annulus floor

241-AZ-101 pump removal trough analysis

International Nuclear Information System (INIS)

Coverdell, B.L.

1995-01-01

As part of the current Hanford mission of environmental cleanup, various long length equipment must be removed from highly radioactive waste tanks. The removal of equipment will utilize portions of the Equipment Removal System for Project W320 (ERS-W320), specifically the 50 ton hydraulic trailer system. Because the ERS-W320 system was designed to accommodate much heavier equipment it is adequate to support the dead weight of the trough, carriage and related equipment for 241AZ101 pump removal project. However, the ERS-W320 components when combined with the trough and its' related components must also be analyzed for overturning due to wind loads. Two troughs were designed, one for the 20 in. diameter carriage and one for the 36 in. diameter carriage. A proposed 52 in. trough was not designed and, therefore is not included in this document. In order to fit in the ERS-W320 strongback the troughs were design with the same widths. Structurally, the only difference between the two troughs is that more material was removed from the stiffener plates on the 36 in trough. The reduction in stiffener plate material reduces the allowable load. Therefore, only the 36 in. trough was analyzed

Tank 241-TY-101 Tank Characterization Plan

International Nuclear Information System (INIS)

Homi, C.S.

1995-01-01

This document is a plan which serves as the contractual agreement between the Characterization Program, Sampling Operations, Oak Ridge National Laboratory, and PNL tank vapor program. The scope of this plan is to provide guidance for the sampling and analysis of vapor samples from tank 241-TY-101

Tank 241-AZ-102 Privatization Push Mode Core Sampling and Analysis Plan

International Nuclear Information System (INIS)

TEMPLETON, A.M.

1999-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AZ-102

Tank 241-AZ-102 Privatization Push Mode Core Sampling and Analysis Plan

International Nuclear Information System (INIS)

RASMUSSEN, J.H.

2000-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AZ-102

Tank characterization report for double-shell tank 241-AP-101. Revision 1

International Nuclear Information System (INIS)

Conner, J.M.

1997-01-01

One major function of the Tank Waste Remediation System (TWRS) is to characterize wastes m support of waste management and disposal activities at the Hanford Site. Analytical data from sampling and analysis and other available information about a tank are compiled and maintained in a tank characterization report (TCR). This report and its appendixes serve as the TCR for double-shell tank 241-AP-101. The objectives of this report are to use characterization data in response to technical issues associated with tank 241-AP-101 waste; and to provide a standard characterization of this waste in terms of a best-basis inventory estimate. Section 2.0 summarizes the response to technical issues, Section 3.0 provides the best-basis inventory estimate, and Section 4.0 makes recommendations about safety status and additional sampling needs. The appendixes contain supporting data and information. This report supported the requirements of the Hanford Federal Facility Agreement and Consent Order, Milestone M-44-05. The characterization information in this report originated from sample analyses and known historical sources. Appendix A provides historical information for tank 241-AP-101 including surveillance, information, records pertaining to waste transfers and tank operations, and expected tank contents derived from a model based upon process knowledge. Appendix B summarizes recent sampling events and historical sampling information. Tank 241-AP-101 was grab sampled in November 1995, when the tank contained 2,790 kL (737 kgal) of waste. An addition1034al 1,438 kL (380 kgal) of waste was received from tank 241-AW-106 in transfers on March 1996 and January 1997. This waste was the product of the 242-A Evaporator Campaign 95-1. Characterization information for the additional 1,438 kL (380 kgal) was obtained using grab sampling data from tank 241-AW-106 and a slurry sample from the evaporator. Appendix C reports on the statistical analysis and numerical manipulation of data used in

Hanford Double-Shell Tank Extent-of-Condition Construction Review

International Nuclear Information System (INIS)

Venetz, Theodore J.; Johnson, Jeremy M.; Gunter, Jason R.; Barnes, Travis J.; Washenfelder, Dennis J.; Boomer, Kayle D.

2013-01-01

During routine visual inspections of Hanford double-shell waste tank 241-AY-102 (AY-102), anomalies were identified on the annulus floor which resulted in further evaluations. Following a formal leak assessment in October 2012, Washington River Protection Solutions, LLC (WRPS) determined that the primary tank of AY-102 was leaking. The formal leak assessment, documented in RPP-ASMT-53793,Tank 241-AY-102 Leak Assessment Report, identified first-of-a-kind construction difficulties and trial-and-error repairs as major contributing factors to tank failure. To determine if improvements in double-shell tank (DST) construction occurred after construction of tank AY-102, a detailed review and evaluation of historical construction records were performed for the first three DST tank farms constructed, which included tanks 241-AY-101, 241-AZ-101, 241-AZ-102, 241-SY-101, 241-SY-102, and 241-SY-103. The review for these six tanks involved research and review of dozens of boxes of historical project documentation. These reviews form a basis to better understand the current condition of the three oldest Hanford DST farms. They provide a basis for changes to the current tank inspection program and also provide valuable insight into future tank use decisions. If new tanks are constructed in the future, these reviews provide valuable 'lessons-learned' information about expected difficulties as well as construction practices and techniques that are likely to be successful

Status of tank 241-SY-101 data analyses

International Nuclear Information System (INIS)

Anantatmula, R.P.

1992-09-01

The Waste Tank Flammable Gas Stabilization Program was established in 1990 to provide for resolution of a major safety issue identified for 23 of the high-level waste tanks at the Hanford Site. The safety issue involves the production, accumulation, and periodic release from these tanks of flammable gases in concentrations exceeding the lower flammability limits. This document deals primarily with tank 241-SY-101 from the SY Tank Farm. The flammable gas condition has existed for this tank since the tank was first filled in the time period from 1977 to 1980. During a general review of waste tank chemical stability in 1988--1989, this situation was re-examined and, in March 1990, the condition was declared to be an unreviewed safety question. Tank 241-SY-101 was placed under special operating restrictions, and a program of investigation was begun to evaluate the condition and determine appropriate courses of action. This report summarizes the data that have become available on tank 241-SY-101 since it was declared as an unreviewed safety question and updates the information reported in an earlier document (WHC-EP-0517). The report provides a technical basis for use in the evaluation of safety risks of the tank and subsequent resolution of the unreviewed safety question

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

Evaluation of 241-AZ tank farm supporting phase 1 privatization waste feed delivery

Energy Technology Data Exchange (ETDEWEB)

CARLSON, A.B.

1998-11-19

This evaluation is one in a series of evaluations determining the process needs and assessing the adequacy of existing and planned equipment in meeting those needs at various double-shell tank farms in support of Phase 1 privatization. A number of tank-to-tank transfers and waste preparation activities are needed to process and feed waste to the private contractor in support of Phase 1 privatization. The scope of this evaluation is limited to process needs associated with 241-AZ tank farm during the Phase 1 privatization.

Evaluation of 241-AZ tank farm supporting phase 1 privatization waste feed delivery

International Nuclear Information System (INIS)

CARLSON, A.B.

1998-01-01

This evaluation is one in a series of evaluations determining the process needs and assessing the adequacy of existing and planned equipment in meeting those needs at various double-shell tank farms in support of Phase 1 privatization. A number of tank-to-tank transfers and waste preparation activities are needed to process and feed waste to the private contractor in support of Phase 1 privatization. The scope of this evaluation is limited to process needs associated with 241-AZ tank farm during the Phase 1 privatization

A survey of available information on gas generation in tank 241-SY-101: Hanford Tank Safety Project

International Nuclear Information System (INIS)

Strachan, D.M.; Reynolds, D.A.; Siemer, D.D.; Wallace, R.W.

1991-03-01

As a result of a concerted effort to determine the chemical and physical mechanisms underlying the generation and episodic release of gases from tank 241-SY-101, more commonly known as tank 101-SY, the Tank Waste Science Panel has been established at the Pacific Northwest Laboratory. Four of the members of this panel met to screen the available information on tank 101-SY and provide to the remaining members a shortened list of references that could be used to assess the mechanisms underlying the generation and episodic release of gases from tank 101-SY. This document is the result of this preliminary screening of information for the Tank Waste Science Panel and was provided to the Panel members at their first meeting. 14 refs., 3 tabs

Tank 241-TY-101 vapor sampling and analysis tank characterization report

International Nuclear Information System (INIS)

Huckaby, J.L.

1995-01-01

This report presents the details of the Hanford waste tank characterization study for tank 241-TY-101. The drivers and objectives of the headspace vapor sampling and analysis were in accordance with procedure that were presented in other reports. The vapor and headspace gas samples were collected and analyzed to determine the potential risks to tank farm workers due to fugitive emissions from the tank

Tank characterization report for single-shell tank 241-A-101

International Nuclear Information System (INIS)

Field, J.M.

1997-01-01

This document summarizes the information on the historical uses, present status, and the sampling and analysis results of waste stored in Tank 241-A-101. This tank has been listed on the Hydrogen Watch List. This report supports the requirements of the Tri-Party Agreement Milestone M-44-10

Tank SY-101 void fraction instrument functional design criteria

International Nuclear Information System (INIS)

McWethy, L.M.

1994-01-01

This document presents the functional design criteria for design, analysis, fabrication, testing, and installation of a void fraction instrument for Tank SY-101. This instrument will measure the void fraction in the waste in Tank SY-101 at various elevations

Tank 241-C-101 tank characterization plan

International Nuclear Information System (INIS)

Schreiber, R.D.

1994-01-01

This document is a plan which serves as the contractual agreement between the Characterization Program, Sampling Operations, WHC 222-S Laboratory, and PNL 325 Analytical Chemistry Laboratory. The scope of this plan is to provide guidance for the sampling and analysis of samples from tank 241-C-101

Ultrasonic Examination of Double-Shell Tank 241-AZ-102 Examination Completed August 2003

International Nuclear Information System (INIS)

Pardini, Allan F.; Posakony, Gerald J.

2003-01-01

COGEMA Engineering Corporation (COGEMA), under a contract from CH2M Hill Hanford Group (CH2M Hill), has performed an ultrasonic nondestructive examination of selected portions of Double-Shell Tank 241-AZ-102. The purpose of this examination was to provide information that could be used to evaluate the integrity of the wall of the primary tank. The requirements for the ultrasonic examination of Tank 241-AZ-102 were to detect, characterize (identify, size, and locate), and record measurements made of any wall thinning, pitting, or cracks that might be present in the wall of the primary tank. Any measurements that exceed the requirements set forth in the Engineering Task Plat (ETP), RPP-11832 (Jensen 2002) and summarized on page 1 of this document, are reported to CH2M Hill and the Pacific Northwest National Laboratory (PNNL) for further evaluation. Under the contract with CH2M Hill, all data is to be recorded on disk and paper copies of all measurements are provided to PNNL for third-party evaluation. PNNL is responsible for preparing a report that describes the results of the COGEMA ultrasonic examinations

Investigation of Tank 241-AN-101 Floating Solids

Energy Technology Data Exchange (ETDEWEB)

Kraft, Douglas P. [Washington River Protection Solutions, LLC, Richland, VA (United States); Meznarich, H. K. [Washington River Protection Solutions, LLC, Richland, VA (United States)

2017-10-30

Tank 241-AN-101 is the receiver tank for retrieval of several C-Farms waste tanks, including Tanks 241-C-102 and 241-C-111. Tank 241 C 111 received first-cycle decontamination waste from the bismuth phosphate process and Plutonium and Uranium Extraction cladding waste, as well as hydraulic fluid. Three grab samples, 1AN-16-01, 1AN-16-01A, and 1AN-16-01B, were collected at the surface of Tank 241-AN-101 on April 25, 2016, after Tank 241-C-111 retrieval was completed. Floating solids were observed in the three grab samples in the 11A hot cell after the samples were received at the 222-S Laboratory. Routine chemical analyses, solid phase characterization on the floating and settled solids, semivolatile organic analysis mainly on the aqueous phase for identification of degradation products of hydraulic fluids were performed. Investigation of the floating solids is reported.

Test plan for Tank 241-AW-101 solubility screening tests

International Nuclear Information System (INIS)

Person, J.C.

1998-01-01

Tank 241-AW-101 (101-AW) has been identified as one of the early tanks to be for retrieved for low level waste pretreatment and immobilization and retrieval of the tank waste may require dilution. This test is to determine the effects of dilution on the mass of solids and their composition. This test plan gives test instructions, example data sheets, a waste compatibility review, and a waste stream fact sheet. This test Plan is similar to tests on tanks 241-AN-102 (Person 1998a) and 241-AN-107 (Person 1998 b). The 101-AW tests will be done with composites of liquid and solids from grab samples that were taken in 1998 (Benar 1998). Future revisions of the Tank Sampling and Analysis Plan (Benar 1998) may change the details of the work performed under this test plan

Chemical compatibility of tank wastes in tanks 241-C-106, 241-AY-101, and 241-AY-102

International Nuclear Information System (INIS)

Sederburg, J.P.

1994-01-01

This report documents the chemical compatibility of waste types within tanks 241-C-106, 241-AY-101, and 241-AY-102. This information was compiled to facilitate the transfer of tank 241-C-106 waste to tank 241-AY-102 utilizing supernatant from tank 241-AY-101 as the sluicing medium. This document justifies that no chemical compatibility safety issues currently understood, or theorized from thermodynamic modeling, will result from the intended sluice transfer operation

Engineering test plan for Tank 241-SY-101 in situ viscometer

International Nuclear Information System (INIS)

Sobocinski, R.G.; Stokes, T.I.; Pearce, K.L.

1994-11-01

To obtain in situ measurements of the rheological properties within tank 241-SY-101, this document will implement the test strategy defined in PNLMIT-041994, acquisition and Reduction of Data Obtained in Tank SY-101 with the Ball Rheometer. Instructions for all sequences are defined within the procedure. All safety requirements as defined in LA-UR-92-3196, A Safety Assessment for Proposed Mixing Operations to Mitigate Episodic Gas Releases in Tank 241-101-SY have been implemented into this procedure

Criticality safety analysis of Hanford Waste Tank 241-101-SY

International Nuclear Information System (INIS)

Perry, R.T.; Sapir, J.L.; Krohn, B.J.

1993-01-01

As part of a safety assessment for proposed pump mixing operations to mitigate episodic gas releases in Tank 241-101-SY at the Hanford Site, Richland, Washington, a criticality safety analysis was made using the Sn transport code ONEDANT. The tank contains approximately one million gallons of waste and an estimated 910 G of plutonium. the criticality analysis considers reconfiguration and underestimation of plutonium content. The results indicate that Tank SY-101 does not present a criticality hazard. These methods are also used in criticality analyses of other Hanford tanks

PSA results for Hanford high level waste Tank 101-SY

Energy Technology Data Exchange (ETDEWEB)

MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W. [Los Alamos National Lab., NM (United States); Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J. [PLG, Inc., Newport Beach, CA (United States)

1993-10-01

Los Alamos National Laboratory has performed a comprehensive probabilistic safety assessment (PSA) that includes consideration of external events for the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases (``burps``) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is underway in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. This PSA for Tank 101-SY, which did not consider the pump experiment or future tank-remediation activities, involved three distinct tasks. First, the accident sequence analysis identified and quantified those potential accidents whose consequences result in tank material release. Second, characteristics and release paths for the airborne and liquid radioactive source terms were determined. Finally, the consequences, primarily onsite and offsite potential health effects resulting from radionuclide release, were estimated, and overall risk curves were constructed. An overview of each of these tasks and a summary of the overall results of the analysis are presented in the following sections.

PSA results for Hanford high level waste Tank 101-SY

International Nuclear Information System (INIS)

MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W.; Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J.

1993-01-01

Los Alamos National Laboratory has performed a comprehensive probabilistic safety assessment (PSA) that includes consideration of external events for the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases (''burps'') a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is underway in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. This PSA for Tank 101-SY, which did not consider the pump experiment or future tank-remediation activities, involved three distinct tasks. First, the accident sequence analysis identified and quantified those potential accidents whose consequences result in tank material release. Second, characteristics and release paths for the airborne and liquid radioactive source terms were determined. Finally, the consequences, primarily onsite and offsite potential health effects resulting from radionuclide release, were estimated, and overall risk curves were constructed. An overview of each of these tasks and a summary of the overall results of the analysis are presented in the following sections

Tank characterization report for single-shell tank 241-B-101

International Nuclear Information System (INIS)

Schreiber, R.D.

1996-01-01

This document summarizes the information on the historical uses, present status, and the sampling and analysis results of waste stored in tank 241-B-101. This report supports the requirements of Tri-Party Agreement Milestone M-44-09

Tank 241-SY-101 push mode core sampling and analysis plan

International Nuclear Information System (INIS)

CONNER, J.M.

1998-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for push mode core samples from tank 241-SY-101 (SY-101). It is written in accordance with Data Quality Objective to Support Resolution of the Flammable Gas Safety Issue (Bauer 1998), Low Activity Waste Feed Data Quality Objectives (Wiemers and Miller 1997 and DOE 1998), Data Quality Objectives for TWRS Privatization Phase I: Confirm Tank T is an Appropriate Feed Source for Low-Activity Waste Feed Batch X (Certa 1998), and the Tank Safety Screening Data Quality Objective (Dukelow et al. 1995). The Tank Characterization Technical Sampling Basis document (Brown et al. 1998) indicates that these issues apply to tank SY-101 for this sampling event. Brown et al. also identifies high-level waste, regulatory, pretreatment and disposal issues as applicable issues for this tank. However, these issues will not be addressed via this sampling event

Mixer pump long term operations plan for Tank 241-SY-101 mitigation

International Nuclear Information System (INIS)

Irwin, J.J.

1994-01-01

This document provides the general Operations Plan for performance of the mixer pump long term operations for Tank 241-SY-101 mitigation of gas retention and periodic release in Tank 101-SY. This operations plan will utilize a 112 kW (150 hp) mixing pump to agitate/suspend the particulates in the tank

Test Results for Caustic Demand Measurements on Tank 241-AX-101 and Tank 241-AX-103 Archive Samples

International Nuclear Information System (INIS)

Doll, Stephanie R.; Bolling, Stacie D.

2016-01-01

Caustic demand testing is used to determine the necessary amount of caustic required to neutralize species present in the Hanford tank waste and obtain a target molarity of free hydroxide for tank corrosion control. The presence and quantity of hydroxide-consuming analytes are just as important in determining the caustic demand as is the amount of free hydroxide present. No single data point can accurately predict whether a satisfactory hydroxide level is being met, as it is dependent on multiple factors (e.g., free hydroxide, buffers, amphoteric metal hydroxides, bicarbonate, etc.). This enclosure contains the caustic demand, scanning electron microscopy (SEM), polarized light microscopy (PLM), and X-ray diffraction (XRD) analysis for the tank 241-AX-101 (AX-101) and 241-AX-103 (AX-103) samples. The work was completed to fulfill a customer request outlined in the test plan, WRPS-1505529, ''Test Plan and Procedure for Caustic Demand Testing on Tank 241-AX-101 and Tank 241-AX-103 Archive Samples.'' The work results will provide a baseline to support planned retrieval of AX-101 and AX-103.

Test Results for Caustic Demand Measurements on Tank 241-AX-101 and Tank 241-AX-103 Archive Samples

Energy Technology Data Exchange (ETDEWEB)

Doll, Stephanie R. [Washington River Protection Solutions, Richland, WA (United States); Bolling, Stacie D. [Washington River Protection Solutions, Richland, WA (United States)

2016-07-14

Caustic demand testing is used to determine the necessary amount of caustic required to neutralize species present in the Hanford tank waste and obtain a target molarity of free hydroxide for tank corrosion control. The presence and quantity of hydroxide-consuming analytes are just as important in determining the caustic demand as is the amount of free hydroxide present. No single data point can accurately predict whether a satisfactory hydroxide level is being met, as it is dependent on multiple factors (e.g., free hydroxide, buffers, amphoteric metal hydroxides, bicarbonate, etc.). This enclosure contains the caustic demand, scanning electron microscopy (SEM), polarized light microscopy (PLM), and X-ray diffraction (XRD) analysis for the tank 241-AX-101 (AX-101) and 241-AX-103 (AX-103) samples. The work was completed to fulfill a customer request outlined in the test plan, WRPS-1505529, “Test Plan and Procedure for Caustic Demand Testing on Tank 241-AX-101 and Tank 241-AX-103 Archive Samples.” The work results will provide a baseline to support planned retrieval of AX-101 and AX-103.

Process Control Plan for Tank 241-SY-101 Surface Level Rise Remediation

International Nuclear Information System (INIS)

ESTEY, S.D.

1999-01-01

The tank 241-SY-101 transfer system was conceived and designed to address the immediate needs presented by rapidly changing waste conditions in tank 241-SY-101. Within the last year or so, the waste in this tank has exhibited unexpected behavior (Rassat et al. 1999) in the form of rapidly increasing crust growth. This growth has been brought about by a rapidly increasing rate of gas entrapment within the crust. It has been conceived that the lack of crust agitation beginning upon the advent of mixer pump operations may have set-up a more consolidated, gas impermeable barrier when compared to a crust regularly broken up by the prior buoyant displacement events within the tank. As a result, a series of level-growth remediation activities have been developed for tank 241-SY-101. The initial activities are also known as near-term crust mitigation. The first activity of near-term mitigation is to perform the small transfer of convective waste from tank 241-SY-101 into tank 241-SY-102. A 100 kgal transfer represents about a 10% volume reduction allowing a 10% water in-tank dilution. Current thinking holds that this should be enough to dissolve nitrite solids in the crust and perhaps largely eliminate gas retention problem in the crust (Raymond 1999). Additional mitigation activities are also planned on less constrained schedules. The net affect of the small transfer and follow-on mitigation activities for tank 241-SY-101 is strongly believed to be the remediation of tank 241-SY-101 as a flammable gas safety concern. The process for remediating the tank will require two or more transfer/dilution cycles. In-tank dilution will begin shortly after the initial transfer and the total dilution required to reach the final state is estimated to be between 250 to 400K gallons. The final state of the waste will not require any active measures to safely store the waste and operation of the mixer pump will no longer be necessary. The remediation activities are centered on a purpose

EFFECTS OF CHEMISTRY AND OTHER VARIABLES ON CORROSION AND STRESS CORROSION CRACKING IN HANFORD DOUBLE SHELL TANKS

Energy Technology Data Exchange (ETDEWEB)

BROWN MH

2008-11-13

Laboratory testing was performed to develop a comprehensive understanding of the corrosivity of the tank wastes stored in Double-Shell Tanks using simulants primarily from Tanks 241-AP-105, 241-SY-103 and 241-AW-105. Additional tests were conducted using simulants of the waste stored in 241-AZ-102, 241-SY-101, 241-AN-107, and 241-AY-101. This test program placed particular emphasis on defining the range of tank waste chemistries that do not induce the onset of localized forms of corrosion, particularly pitting and stress corrosion cracking. This document summarizes the key findings of the research program.

EFFECTS OF CHEMISTRY AND OTHER VARIABLES ON CORROSION AND STRESS CORROSION CRACKING IN HANFORD DOUBLE-SHELL TANKS

International Nuclear Information System (INIS)

Brown, M.H.

2008-01-01

Laboratory testing was performed to develop a comprehensive understanding of the corrosivity of the tank wastes stored in Double-Shell Tanks using simulants primarily from Tanks 241-AP-105, 241-SY-103 and 241-AW-105. Additional tests were conducted using simulants of the waste stored in 241-AZ-102, 241-SY-101, 241-AN-107, and 241-AY-101. This test program placed particular emphasis on defining the range of tank waste chemistries that do not induce the onset of localized forms of corrosion, particularly pitting and stress corrosion cracking. This document summarizes the key findings of the research program

Preliminary tank characterization report for single-shell tank 241-TX-101: best-basis inventory

International Nuclear Information System (INIS)

Kupfer, M.J.

1997-01-01

This document is a preliminary Tank Characterization Report (TCR). It only contains the current best-basis inventory (Appendix D) for single-shell tank 241-TX-101. No TCRs have been previously issued for this tank, and current core sample analyses are not available. The best-basis inventory, therefore, is based on an engineering assessment of waste type, process flowsheet data, early sample data, and/or other available information. The Standard Inventories of Chemicals and Radionuclides in Hanford Site Tank Wastes describes standard methodology used to derive the tank-by-tank best-basis inventories. This preliminary TCR will be updated using this same methodology when additional data on tank contents become available

Tank 241-AZ-102 Privatization Push Mode Core Sampling and Analysis Plan; FINAL

International Nuclear Information System (INIS)

TEMPLETON, A.M.

1999-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AZ-102. The purpose of this sampling event is to obtain information about the characteristics of the contents of 241-AZ-102. Push mode core samples will be obtained from risers 15C and 24A to provide sufficient material for the chemical analyses and tests required to satisfy these data quality objectives. The 222-S Laboratory will extrude core samples, composite the liquids and solids, perform chemical analyses, and provide subsamples to the Process Chemistry Laboratory. The Process Chemistry Laboratory will prepare test plans and perform process tests to evaluate the behavior of the 241-AZ-102 waste undergoing the retrieval and treatment scenarios defined in the applicable DQOs. Requirements for analyses of samples originating in the process tests will be documented in the corresponding test plan

Structural analysis and evaluation of the 241SY101 tank annulus heat-up

International Nuclear Information System (INIS)

Ziada, H.H.

1994-01-01

This document provides the structural analysis (static and thermal loads) of the 241SY101 tank to determine the maximum allowable temperature and rate of heating that could be applied to tank 241SY101 through annulus air heating without detrimental effects to the structural integrity of the concrete and steel liner of the tank

Software configuration management plan, 241-AY and 241-AZ tank farm MICON automation system

International Nuclear Information System (INIS)

Hill, L.F.

1997-01-01

This document establishes a Computer Software Configuration Management Plan (CSCM) for controlling software for the MICON Distributed Control System (DCS) located at the 241-AY and 241-AZ Aging Waste Tank Farm facilities in the 200 East Area. The MICON DCS software controls and monitors the instrumentation and equipment associated with plant systems and processes. A CSCM identifies and defines the configuration items in a system (section 3.1), controls the release and change of these items throughout the system life cycle (section 3.2), records and reports the status of configuration items and change requests (section 3.3), and verifies the completeness and correctness of the items (section 3.4). All software development before initial release, or before software is baselined, is considered developmental. This plan does not apply to developmental software. This plan applies to software that has been baselined and released. The MICON software will monitor and control the related instrumentation and equipment of the 241-AY and 241-AZ Tank Farm ventilation systems. Eventually, this software may also assume the monitoring and control of the tank sludge washing equipment and other systems as they are brought on line. This plan applies to the System Cognizant Manager and MICON Cognizant Engineer (who is also referred to herein as the system administrator) responsible for the software/hardware and administration of the MICON system. This document also applies to any other organizations within Tank Farms which are currently active on the system including system cognizant engineers, nuclear operators, technicians, and control room supervisors

Engineering evaluation of alternatives: Managing the assumed leak from single-shell Tank 241-T-101

International Nuclear Information System (INIS)

Brevick, C.H.; Jenkins, C.

1996-02-01

At mid-year 1992, the liquid level gage for Tank 241-T-101 indicated that 6,000 to 9,000 gal had leaked. Because of the liquid level anomaly, Tank 241-T-101 was declared an assumed leaker on October 4, 1992. SSTs liquid level gages have been historically unreliable. False readings can occur because of instrument failures, floating salt cake, and salt encrustation. Gages frequently self-correct and tanks show no indication of leak. Tank levels cannot be visually inspected and verified because of high radiation fields. The gage in Tank 241-T-101 has largely corrected itself since the mid-year 1992 reading. Therefore, doubt exists that a leak has occurred, or that the magnitude of the leak poses any immediate environmental threat. While reluctance exists to use valuable DST space unnecessarily, there is a large safety and economic incentive to prevent or mitigate release of tank liquid waste into the surrounding environment. During the assessment of the significance of the Tank 241-T-101 liquid level gage readings, Washington State Department of Ecology determined that Westinghouse Hanford Company was not in compliance with regulatory requirements, and directed transfer of the Tank 241-T-101 liquid contents into a DST. Meanwhile, DOE directed WHC to examine reasonable alternatives/options for safe interim management of Tank 241-T-101 wastes before taking action. The five alternatives that could be used to manage waste from a leaking SST are: (1) No-Action, (2) In-Tank Stabilization, (3) External Tank Stabilization, (4) Liquid Retrieval, and (5) Total Retrieval. The findings of these examinations are reported in this study

Steady-State Flammable Gas Release Rate Calculation And Lower Flammability Level Evaluation For Hanford Tank Waste

International Nuclear Information System (INIS)

Meacham, J.E.

2009-01-01

This report assesses the steady state flammability level under off normal ventilation conditions in the tank headspace for 28 double-shell tanks (DST) and 149 single shell-tanks (SST) at the Hanford Site. Flammability was calculated using estimated gas release rates, Le Chatelier's rule, and lower flammability limits of fuels in an air mixture. This revision updates the hydrogen generation rate input data for all 177 tanks using waste composition information from the Best Basis Inventory Detail Report (data effective as of August 4,2008). Assuming only barometric breathing, the shortest time to reach 25% of the lower flammability limit is 11 days for DSTs (i.e., tank 241-AZ-10l) and 36 days for SSTs (i.e., tank 241-B-203). Assuming zero ventilation, the shortest time to reach 25% of the lower flammability limit is 10 days for DSTs (i.e., tank 241-AZ-101) and 34 days for SSTs (i.e., tank 241-B-203).

Thermal modeling of tanks 241-AW-101 and 241-AN-104 with the TEMPEST code

International Nuclear Information System (INIS)

Antoniak, Z.I.; Recknagle, K.P.

1995-07-01

The TEMPEST code was exercised in a preliminary study of double-shell Tanks 241 -AW-101 and 241-AN-104 thermal behavior. The two-dimensional model used is derived from our earlier studies on heat transfer from Tank 241-SY-101. Several changes were made to the model to simulate the waste and conditions in 241-AW-101 and 241-AN-104. The nonconvective waste layer was assumed to be 254 cm (100 in.) thick for Tank 241-AW-101, and 381 cm (150 in.) in Tank 241-AN-104. The remaining waste was assumed, for each tank, to consist of a convective layer with a 7.6-cm (3-inch) crust on top. The waste heat loads for 241-AW-101 and 241-AN-104 were taken to be 10 kW (3.4E4 Btu/hr) and 12 kW (4.0E4 Btu/hr), respectively. Present model predictions of maximum and convecting waste temperatures are within 1.7 degrees C (3 degrees F) of those measured in Tanks 241-AW-101 and 241-AN-104. The difference between the predicted and measured temperature is comparable to the uncertainty of the measurement equipment. These models, therefore, are suitable for estimating the temperatures within the tanks in the event of changing air flows, waste levels, and/or waste configurations

Independent Review of Tank 241-AY-101 Fitness for Service

International Nuclear Information System (INIS)

Stewart, Charles W.; Bush, Spencer H.; Delegard, Calvin H.; Elmore, Monte R.; Johnson, A Burton Jr.; Pardini, Allan F.; Posakony, Gerald J.; Simonen, Fredric A.; Terry, Michael T.; Zapp, Philip E.

2003-01-01

Video inspections in the annulus of Hanford double-shell waste storage tank 241-AY-101 in 2001 and earlier showed rust over large areas of the primary and secondary tank walls. These observations led to extensive on-destructive inspections and analyses to determine the extent and severity of the corrosion and correction of several operational deficiencies that contributed to the problem. PNNL has performed an independent review of the accumulated evidence from these efforts to determine whether unacceptable conditions were corrected and the current condition of the tank meets or exceeds technical and operational requirements. The findings of this review are the subject of this report. The conclusion is that AY-101 is fit for service without restriction subject to several recommendations for further inspections and analyses

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

Nuclear waste inventory characterization for mixer pumps and long length equipment removed from Hanford waste tanks

International Nuclear Information System (INIS)

Troyer, G.L.

1998-01-01

The removal and disposition of contaminated equipment from Hanford high-level nuclear waste tanks presents many challenges. One of which is the characterization of radioactive contaminants on components after removal. A defensible assessment of the radionuclide inventory of the components is required for disposal packaging and classification. As examples of this process, this paper discusses two projects: the withdrawal of thermocouple instrument tubes from Tank 101-AZ, and preparation for eventual replacement of the hydrogen mitigation mixer pump in Tank 101-SY. Emphasis is on the shielding analysis that supported the design of radiation detection systems and the interpolation of data recorded during the equipment retrieval operations

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

Hanford's Battle with Nuclear Waste Tank SY-101: Bubbles, Toils, and Troubles

International Nuclear Information System (INIS)

Stewart, Charles W.

2006-01-01

Radioactive waste tank SY-101 is one of 177 big underground tanks that store waste from decades of plutonium production at the Hanford Nuclear Reservation in central Washington State. The chemical reactions and radioactivity in all the tanks make bubbles of flammable gas, mainly hydrogen along with a little methane and ammonia. But SY-101 was the most potent gas producer of all. Every few months the gas built up in the million gallons of extra-thick slurry until it suddenly came up in great rushing ''burps''. A few of the tank's larger burps let off enough gas to make the air space at the top of the tank flammable for a few hours. This flammable gas hazard became a dominating force in DOE nuclear waste management politics in the last two decades of the 20th century. It demanded the toil of scientists, managers, and officials from the time it was filled in 1980, until it was finally declared safe in January 2001. The tank seemed almost a personality--acting with violence and apparent malice, hiding information about itself, deceiving us with false indications, and sometimes lulling us into complacency only to attack in a new way. From 1990 through 1993, SY-101's flammable gas troubles were acknowledged as the highest priority safety issue in the entire DOE complex. Uncontrolled crust growth demanded another high-priority remedial effort from 1998 through April 2000. The direct cost of the bubbles, toils, and troubles was high. Overall, the price of dealing with the real and imagined hazards in SY-101 may have reached $250 million. The indirect cost was also high. Spending all this money fighting SY-101?s safety issues only stirred radioactive waste up and moved it around, but accomplished no cleanup whatever. Worse yet, the flammable gas problem spawned suspicions of a much wider danger that impeded and complicated cleanup in other 176 waste tanks for a decade. The real cleanup job has yet to be done. The SY-101 story is really about the collective experience of

Hanford Site Tank 241-SY-101, damaged equipment removal

International Nuclear Information System (INIS)

Titzler, P.A.; Legare, D.E.; Barrus, H.G.

1993-11-01

Hanford Site Tank 241-SY-101 has a history of generating hydrogen-nitrous oxide gases. The gases are generated and trapped in the non-convective waste layer near the bottom of the 23-m- (75-ft-) diameter underground tank. Approximately every three months the pressure in the tank is relieved as the trapped gases are released through or around the surface crust into the tank dome. This process moves large amounts of liquid waste and crust material around in the tank. The moving waste displaced air lances and thermocouple assemblies (2-in. schedule-40 pipe) installed in four tank risers and permanently bent them to a maximum angle of 40 degrees. The bends were so severe that assemblies could not be removed from the tank using the originally designed hardware. Just after the tank releases the trapped gas, a 20-to-30-day work ''window'' opens

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

Final Report DM1200 Tests With AZ 101 HLW Simulants VSL-03R3800-4, Rev. 0, 2/17/04

International Nuclear Information System (INIS)

Kruger, A.A.; Matlack, K.S.; Bardakci, T.; D'Angelo, N.A.; Gong, W.; Kot, W.K.; Pegg, I.L.

2011-01-01

This report documents melter and off-gas performance results obtained on the DM 1200 HLW Pilot Melter during processing of simulated HLW AZ-101 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW AZ-101 feed; determine the effect of bubbling rate and feed solids content on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post-test inspections of system components. The test objectives (including test success criteria), along with how they were met, are outlined in a table.

FINAL REPORT DM1200 TESTS WITH AZ 101 HLW SIMULANTS VSL-03R3800-4 REV 0 2/17/04

Energy Technology Data Exchange (ETDEWEB)

KRUGER AA; MATLACK KS; BARDAKCI T; D' ANGELO NA; GONG W; KOT WK; PEGG IL

2011-12-29

This report documents melter and off-gas performance results obtained on the DM 1200 HLW Pilot Melter during processing of simulated HLW AZ-101 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW AZ-101 feed; determine the effect of bubbling rate and feed solids content on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post-test inspections of system components. The test objectives (including test success criteria), along with how they were met, are outlined in a table.

An assessment of the dilution required to mitigate Hanford tank 241-SY-101

International Nuclear Information System (INIS)

Hudson, J.D.; Bredt, P.R.; Felmy, A.R.; Stewart, C.W.; Tingey, J.M.; Trent, D.S.; Barney, G.S.; Herting, D.L.; Larrick, A.P.; Reynolds, D.A.

1995-02-01

A group of experts from PNL and WHC convened November 2 and 3, 1994, to screen the current state of knowledge about dilution and reach a consensus on the minimum dilution ratio that will achieve passive mitigation of Tank 241-SY-101 wastes and the dilution ratio that would satisfy the given cross-site transfer criteria with reasonable assurance. The panel evaluated the effects of dilution on the parameters important in gas generation, retention, and release and reached the following conclusions, which are deduced from the existing body of data, experience, and analyses: (1) Dissolution of solids is the single most important aspect of mitigation by dilution. We are confident that diluting until nitrates, nitrites, and aluminum salts are dissolved will mitigate Hanford flammable gas tanks; (2) Sufficient solids dissolution can be achieved in Tank 241-SY-101 at a dilution ratio of 1:1, which will result in a average specific gravity of approximately 1.35. It is likely that a 0.5:1 dilution will also mitigate 241-SY-101, but the current uncertainty is too high to recommend this dilution ratio; (3) The recommended dilution requires a diluent with at least 2 molar free hydroxide, because aluminum probably precipitates at lower hydroxide concentrations. The transfer criteria for Tank 241-SY-101 waste were also evaluated. These criteria have been specified as solids content ≤30% (volume), viscosity ≤30% cP and density <1.5 g/mL. (1) Solids content is the limiting criterion if it is defined as volume fraction of settled solids. A 1:1 dilution will satisfy this criterion at nominal premixing conditions in Tank 241-SY-101; however, analysis of Window E core samples suggests that up to 1.5:1 might be required. If the solids content is interpreted simply as solids volume fraction no further dilution is necessary, because Tank 241-SY-101 waste (excluding the crust) is already below 30%; (2) Bulk density is the next limiting criterion and is met at 0.4:1 dilution

SAFETY ANALYSIS APPROACH TO TANK 241-SY-101 REMEDIATION ACTIVITIES

International Nuclear Information System (INIS)

RYAN, G.W.

2000-01-01

An Unreviewed Safety Question was declared related to the unexplained waste surface level growth in high-level radioactive waste storage Tank 241-SY-101 at the Hanford Site in Richland, Washington. Because the waste surface level in Tank 241-SY-101 was growing in a manner inconsistent with previous behavior, the following issues of concern were recognized: (1) The continually rising surface level had the potential to reach physical encumbrances or limits within the tank (e.g., instrumentation, cameras, established Authorization Basis limits, and the double containment boundary) and the potential to significantly change the consequences of previously analyzed accidents (e.g., flammable gas deflagrations). (2) The presence of new hazards because of significant quantities of flammable gas retained in the crust (e.g., crust collapse gas-release events). (3) The potential to inhibit information gathering related to the existing hazards in the tank (e.g., unable to determine surface level to assess the potential for large gas releases). In response to this situation, a Contractor Project Team, which included Department of Energy representation, was formed to constructively address the issue. The team was responsible for developing and evaluating remediation options and executing the chosen option for remediating the surface level rise issue for Tank 241-SY-101. From an Authorization Basis perspective, the following important aspects will be discussed in this paper: (1) The integrated nature of the Project Team. The team consisted of all the organizations necessary to ensure that the time available to remediate Tank 241-SY-101 was effectively used. Most notable is the connectivity of the Nuclear Safety and Licensing organization with the Engineering, Design, and Operations organizations. (2) The ability of the safety analysis support to adjust to and address evolving Project Team goals and dynamic tank conditions. (3) Due to the urgency to mitigate this developing issue

Probabilistic safety assessment for Hanford high-level waste tank 241-SY-101

Energy Technology Data Exchange (ETDEWEB)

MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W. [Los Alamos National Lab., NM (United States); Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J. [PLG, Inc., Newport Beach, CA (United States)

1994-05-01

Los Alamos National Laboratory (Los Alamos) is performing a comprehensive probabilistic safety assessment (PSA), which will include consideration of external events for the 18 tank farms at the Hanford Site. This effort is sponsored by the Department of Energy (DOE/EM, EM-36). Even though the methodology described herein will be applied to the entire tank farm, this report focuses only on the risk from the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases ({open_quotes}burps{close_quotes}) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed first because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is being conducted in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. At the Hanford Site there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/salt cake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is about 60 million gal., which contains approximately 120 million Ci of radioactivity.

Probabilistic safety assessment for Hanford high-level waste tank 241-SY-101

International Nuclear Information System (INIS)

MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W.; Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J.

1994-05-01

Los Alamos National Laboratory (Los Alamos) is performing a comprehensive probabilistic safety assessment (PSA), which will include consideration of external events for the 18 tank farms at the Hanford Site. This effort is sponsored by the Department of Energy (DOE/EM, EM-36). Even though the methodology described herein will be applied to the entire tank farm, this report focuses only on the risk from the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases (open-quotes burpsclose quotes) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed first because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is being conducted in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. At the Hanford Site there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/salt cake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is about 60 million gal., which contains approximately 120 million Ci of radioactivity

Acquisition and reduction of data obtained from Tank 101-SY in-situ ball rheometer

International Nuclear Information System (INIS)

Shepard, C.L.; Chieda, M.A.; Kirihara, L.J.

1994-12-01

Development of the ball rheometer to measure rheological properties and density of the waste in Hanford Tank 241-SY-101 will be completed around September 1994. Since the ball rheometer project began, a mixer pump has been installed in this tank, and by all accounts this pump has been very successful at mitigating the flammable gas problem associated with Tank 101-SY. Present plans now call for the use of mixer pumps in several other tanks. The ball rheometer will serve as a diagnostic tool for judging the effectiveness of mixing in Tank 101-SY and others and will be one of few in-situ probes available for diagnostic measurements. The in-situ data collection strategy and the methods of data analysis and reduction are presented in this final report concerning this instrument. It is believed that a generalized Bingham fluid model (Herschel-Bulkley fluid model) may be useful for describing at least some of the waste contained in Tank 101-SY, and data obtained in the tank will initially be reduced using this fluid model. The single largest uncertainty in the determination of the drag force on the ball is the drag force which will be experienced by the cable attached to the ball. This drag can be a substantial fraction of the total drag when the ball is deep within the tank. Careful accounting of the cable drag will be important in the reduction of the data. The data collection strategy allows the determination of the waste fluid rheology both in the undisturbed state and after it has been disturbed by the ball. Fluid density will be measured at regular intervals

Tank Vapor Characterization Project: Vapor space characterization of waste Tank A-101, Results from samples collected on June 8, 1995

International Nuclear Information System (INIS)

Pool, K.H.; Clauss, T.W.; McVeety, B.D.; Evans, J.C.; Thomas, B.L.; Olsen, K.B.; Fruchter, J.S.; Ligotke, M.W.

1995-11-01

This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-A-101 (Tank A-101) at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank-farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest National Laboratory (PNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNL. Analyte concentrations were based on analytical results and, where appropriate, sample volumes provided by WHC. A summary of the results is listed in Table 1. Detailed descriptions of the analytical results appear in the text

Operational experience in mitigating flammable gas releases from Hanford Site Tank 241-SY-101

International Nuclear Information System (INIS)

Lentsch, J.W.; Babad, H.; Kirch, N.W.

1995-01-01

Flammable gases consisting of hydrogen, nitrous oxide, ammonia, and methane are periodically released from Hanford Site waste tank 241-SY-101 at concentrations above the flammable limit. A large mixer pump installed in the tank in 1993 has effectively mitigated this problem by continuously releasing small amounts of the flammable gases at the rate they are generated. Tank 241-SY-101 is also equipped with multiple high-sensitivity gas monitoring systems and level detection systems to measure the quantity of gas that is retained in and released from the waste

The effect of dilution on the gas-retention behavior of Tank 241-SY-101 waste

International Nuclear Information System (INIS)

Bredt, P.R.; Tingey, S.M.; Shade, E.H.

1995-09-01

The effect of dilution on gas retention in waste from Tank 241-SY-101 was investigated. A composite sample was prepared from material collected during the Window ''C'' and Window ''E'' sampling events. The composite contained material from both the convective and nonconvective layer in the proportions existing in the tank. Operation of the mixer pump in Tank 241-SY-101 has homogenized the tank material, and dilution of the current waste would require additional mixing; therefore, no attempt was made to use unhomogenized tank waste to prepare the composite. The composite was diluted with 2 M NaOH at ratios of 0.5:1, 0.75: 1, 1:1, and 3:1 per volume (2 M NaOH:tank waste)

Headspace vapor characterization of Hanford Waste Tank AX-101: Results from samples collected on June 15, 1995. Tank Vapor Characterization Project

International Nuclear Information System (INIS)

Pool, K.H.; Clauss, T.W.; Evans, J.C.; McVeety, B.D.

1996-05-01

This report describes the results of vapor samples taken from the headspace of waste storage tank 241-AX-101 (Tank AX-101) at the Hanford Site in Washington State. Pacific Northwest National Laboratory (PNNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze samples for inorganic and organic analytes collected from the tank headspace and ambient air near the tank. The analytical work was performed by the PNNL Vapor Analytical Laboratory (VAL) under the Tank Vapor Characterization Project. Work performed was based on a sample and analysis plan (SAP) prepared by WHC. The SAP provided job-specific instructions for samples, analyses, and reporting. The SAP for this sample job was open-quotes Vapor Sampling and Analysis Planclose quotes, and the sample job was designated S5028. Samples were collected by WHC on June 15, 1995, using the Vapor Sampling System (VSS), a truck-based sampling method using a heated probe inserted into the tank headspace

Historical trends in tank 241-SY-101 waste temperatures and levels

International Nuclear Information System (INIS)

Antoniak, Z.I.

1993-09-01

The gas release and fluctuating level of the waste in tank 241-SY-101 have prompted more detailed interest in its historical behavior, in hopes of achieving a better understanding of its current status. To examine the historical behavior, essentially all of the tank waste temperature and level data record has been retrieved, examined, and plotted in various ways. To aid in interpreting the data, the depth of the non-convective waste layer was estimated by using a least-squares Chebyshev approximation to the temperatures. This report documents the retrieval critical examination, and graphic presentation of 241-SY-101 temperature and waste level histories. The graphic presentations clearly indicate a tank cooling trend that has become precipitous since late 1991. The plots also clearly show the decreasing frequency of waste gas release events, increasing height of the non-convective layer, and larger level drops per event

Chemical compatibility of tank wastes in 241-C-106, 241-AY-101, and 241-AY-102

International Nuclear Information System (INIS)

Sederburg, J.P.

1994-01-01

This report documents the chemical compatibility of waste types within tanks 241-C-106, 241-AY-101, and 241-AY-102. This information was compiled to facilitate the transfer of tank C-106 waste to tank AY-102 utilizing supernatant from AY-101 as the sluicing medium. This document justifies that no chemical compatibility safety issues currently understood, or theorized from thermodynamic modeling, will result from the intended sluice transfer operation

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

Mixer pump test plan for double-shell tank AZ-101. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Symons, G.A.

1996-02-01

Westinghouse Hanford Company has undertaken the task to develop and demonstrate a method of retrieval for double-shell tank waste. Mixer pumps were chosen as the planned method of retrieval for DSTs, based on engineering technology studies, past experience with hydraulic sluicing at the Hanford Site, and experience with mixer pumps at the Westinghouse Savannah River Site.

Mixer pump test plan for double-shell tank AZ-101. Revision 1

International Nuclear Information System (INIS)

Symons, G.A.

1996-02-01

Westinghouse Hanford Company has undertaken the task to develop and demonstrate a method of retrieval for double-shell tank waste. Mixer pumps were chosen as the planned method of retrieval for DSTs, based on engineering technology studies, past experience with hydraulic sluicing at the Hanford Site, and experience with mixer pumps at the Westinghouse Savannah River Site

Acquisition and reduction of data obtained from tank 101-SY in-situ ball rheometer

International Nuclear Information System (INIS)

Shepard, C.L.; Chieda, M.A.; Kirihara, L.J.; Phillips, J.R.; Shekarriz, A.; Terrones, G.; Abbott, J.; Unal, C.; Pasamehmetoglu, K.O.; Graham, A.

1995-02-01

Development of the ball rheometer to measure rheological properties and density of the waste in Hanford Tank 241-SY-101 will be completed around September 1994. This instrument is expected to provide the first-of-its-kind in-situ measurements of the fluid properties of the waste contained within this tank. A mixer pump has been installed in this tank, and this pump has been very successful at mitigating the flammable gas problem associated with Tank 101-SY. The ball rheometer will serve as a diagnostic tool for judging the effectiveness of mixing in Tank 101-SY and others and will be one of few in-situ probes available for diagnostic measurements. Based on experiments performed at Los Alamos National Laboratory and Pacific Northwest Laboratory, it is believed that a generalized Bingham fluid model (Herschel-Bulkley fluid model) may be useful for describing at least some of the waste contained in Tank 101-SY, and data obtained in the tank will initially be reduced using this fluid model. The single largest uncertainty in the determination of the drag force on the ball is the drag force which will be experienced by the cable attached to the ball. This drag can be a substantial fraction of the total drag when the ball is deep within the tank. It is expected that the fluid properties may be history dependent, thus rheological properties of the undisturbed fluid may be different from the same properties after the fluid has been disturbed by passage of the ball. The data collection strategy allows the determination of the waste fluid rheology both in the undisturbed state and after it has been disturbed by the ball. Unlike the rheological parameters, measurement of density requires no model for its interpretation; however, the effects of yield stress may need to be accounted for. This measurement can be made with fairly good accuracy and may provide the most useful data in determination of mixer pump effectiveness

Neutralized current acid waste consolidation management plan

International Nuclear Information System (INIS)

Powell, W.J.; Brown, R.G.; Galbraith, J.; Jensen, C.; Place, D.E.; Reddick, G.W.; Zuroff, W.; Brothers, A.J.

1996-01-01

The scope of this evaluation is to recommend a management plan for the high-heat tank waste, including neutralized current acid waste (NCAW) in AY and AZ Tank Farms, and tank C-106 waste. The movement of solids, liquids and salt cake in the designated tank farms is included. Decision analysis techniques were used to determine a recommended alternative. The recommended course of action was replacement of a 75-hp mixer pump in tank AY-102 and in-tank concentration of tank AZ-102 supernate. The alternative includes transfer fo tank C-106 sludge to tank AY-102, then transfer to tank AY-102 and tank C-106 sludge to tank AZ-101 using the new 75-hp mixer pump installed in tank AY-102. Tank AZ-101 becomes a storage tank for high-level waste (HLW) sludge, with the capacity to mix and transfer sludge as desired

Safety evaluation for adding water to tank 101-SY

International Nuclear Information System (INIS)

Clinton, R.

1994-01-01

This document provides a new water limit for Tank 241-SY-101. The original limit was set at 9600 gallons. The new limit is now 20,000 gallons. There are various activities that require the use of additional water to the tank. The main activity is the removal of the temporary mixer pump. This requires a large amount of water which will exceed the original limit. Also, other activities such as flushing, adding a viscometer, and adding a void fraction meter requires additional water. The new limit safely incorporates these activities and allows room for more future activities

Composition and quantities of retained gas measured in Hanford waste tanks 241-AW-101 A-101, AN-105, AN-104, and AN-103

International Nuclear Information System (INIS)

Shekarriz, A.; Rector, D.R.; Mahoney, L.A.

1997-03-01

This report provides the results obtained for the first five tanks sampled with the Retained Gas Sampler (RGS): Tanks 241-AW-101, A-101, AN-105, AN-104, and AN-103. The RGS is a modified version of the core sampler used at Hanford. It is designed specifically, in concert with the gas extraction equipment in the hot cell, to capture and extrude a gas-containing waste sample in a hermetically sealed system. The retained gases are then extracted and stored in small gas canisters. The composition of the gases contained in the canisters was measured by mass spectroscopy. The total gas volume was obtained from analysis of the extraction process, as discussed in detail throughout this report. The following are the findings of this research: (1) The RGS is a viable approach for measuring retained gases in double- and single-shell waste tanks at Hanford. (2) Local measurements of void fraction with the RGS agree with the results obtained with the void fraction instrument (VFI) in most cases. (3) In the tanks sampled, more than 16% of the retained gas in the nonconvective layer was nitrogen (N 2 ). The fraction of nitrogen gas was approximately 60% in Tank 241-AW-101. This finding shows that not all the retained gas mixtures are flammable. (4) In the tanks sampled, the ratios of hydrogen to oxidizers were observed to be significantly higher than 1; i.e., these tanks are fuel-rich. Based on these observations, the RGS will be used to sample for retained gases in several single-shell tanks at Hanford. The remaining sections of this summary describe the RGS-findings for the first five tanks tested. The results are described in the order in which the tanks were sampled, to reflect the increasing experience on which RGS methods were based

Preliminary tank characterization report for single-shell tank 241-BY-101: best-basis inventory

International Nuclear Information System (INIS)

Kupfer, M.J.

1997-01-01

An effort is underway to provide waste inventory estimates that will serve as standard characterization source terms for the various waste management activities. As part of this effort, an evaluation of available information for single-shell tank 241-BY-101 was performed, and a best-basis inventory was established. This work follows the methodology that was established by the standard inventory task

Preliminary tank characterization report for single-shell tank 241-TY-101: best-basis inventory

International Nuclear Information System (INIS)

Lambert, S.L.

1997-01-01

An effort is underway to provide waste inventory estimates that will serve as standard characterization source terms for the various waste management activities. As part of this effort, an evaluation of available information for single-shell tank 241-TY-101 was performed, and a best-basis inventory was established. This work follows the methodology that was established by the standard inventory task

Acceptance/operational test procedure 101-AW tank camera purge system and 101-AW video camera system

International Nuclear Information System (INIS)

Castleberry, J.L.

1994-01-01

This procedure will document the satisfactory operation of the 101-AW Tank Camera Purge System (CPS) and the 101-AW Video Camera System. The safety interlock which shuts down all the electronics inside the 101-AW vapor space, during loss of purge pressure, will be in place and tested to ensure reliable performance. This procedure is separated into four sections. Section 6.1 is performed in the 306 building prior to delivery to the 200 East Tank Farms and involves leak checking all fittings on the 101-AW Purge Panel for leakage using a Snoop solution and resolving the leakage. Section 7.1 verifies that PR-1, the regulator which maintains a positive pressure within the volume (cameras and pneumatic lines), is properly set. In addition the green light (PRESSURIZED) (located on the Purge Control Panel) is verified to turn on above 10 in. w.g. and after the time delay (TDR) has timed out. Section 7.2 verifies that the purge cycle functions properly, the red light (PURGE ON) comes on, and that the correct flowrate is obtained to meet the requirements of the National Fire Protection Association. Section 7.3 verifies that the pan and tilt, camera, associated controls and components operate correctly. This section also verifies that the safety interlock system operates correctly during loss of purge pressure. During the loss of purge operation the illumination of the amber light (PURGE FAILED) will be verified

Level maintenance for Tank 101-SY mitigation-by-mixing test. Revision 2

International Nuclear Information System (INIS)

Larsen, D.C.

1994-01-01

The Phase A, Phase B and Full Scale testing portions of the Mitigation-By-Mixing Test have demonstrated the effectiveness of the Mixer Pump to maintain the waste in tank 101-SY in the desired mitigated state. The operation of the 101-SY Mixer Pump for short periods of time results in a controlled release of hydrogen gas in concentrations well below the established safety limits. Additionally, it has been shown that operation of the pump on a regular schedule minimizes the historical generation rate of hydrogen inventory in the waste. Generation of hydrogen inventory is exhibited by waste level growth. The primary objective of this procedure is to maintain the waste level in tank 241-SY-101 within the safe operating range as defined by the Safety Assessment and the Test Plan. The secondary objective is to operate the pump on a schedule that maximizes its useful lifespan and prevents the formation of obstructions in the normal flow path of the pump

The effects of heating and dilution on the rheological and physical properties of Tank 241-SY-101 waste

International Nuclear Information System (INIS)

Tingey, J.M.; Bredt, P.R.; Shade, E.H.

1994-10-01

Of the 177 high-level waste underground storage tanks at the Hanford Site, 25 have been identified as being potentially capable of generating and releasing flammable gas. Tank 241-SY-101 has exhibited periodic releases of gas, and in some cases the gas released has exceeded the lower flammable gas limit. The components of the released gas from Tank 241-SY-101 are hydrogen, nitrous oxide, nitrogen, ammonia, carbon monoxide, and methane. A mitigation strategy that may effectively reduce the retention and release of these gases and the release of flammable gases is dilution coupled with eating of the tank wastes. The purpose of this work was to determine changes in rheological and physical properties caused by heating and dilution of actual 241-SY-101 waste. In May and December 1991, following periodic gas releases, samples of the waste in Tank 241-SY-101 were obtained. Current work quantified the effects of heating coupled with NaOH dilution of a combination of waste samples from Tank 241-SY-101 characteristic of a non-convective layer. The experimental approach and results of this heating and dilution study on Tank 241-SY-101 waste samples are described in Sections 2 and 3, respectively. In Section 3.1, a discussion of the rheological properties of the waste as a function of shearing forces, volume percent dilution, and temperature is presented. In Section 3.2, the physical properties of the waste dilutions are described, including the densities of the slurry, filtered solids, and filtrate; the settling behavior; and the percent filtered solids in the composite sample and each of the composite dilutions. A brief discussion of the results and uncertainties is given is Section 3.3. The conclusions of this investigation are reported in Section 4

Hanford Tanks 241-AY-102 and 241-BX-101: Sludge Composition and Contaminant Release Data

International Nuclear Information System (INIS)

Krupka, Kenneth M.; Deutsch, William J.; Lindberg, Michael J.; Cantrell, Kirk J.; Hess, Nancy J.; Schaef, Herbert T.; Arey, Bruce W.

2004-01-01

This report describes the results of testing sludge samples from Hanford tanks 241-AY-102 (AY-102) and 241-BX-101 (BX-101). These tests were conducted to characterize the sludge and assess the water leachability of contaminants from the solids. This work is being conducted to support the tank closure risk assessments being performed by CH2M HILL Hanford Group, Inc. for the U.S. Department of Energy. This is the first report of testing of BX-101 sludge and the second report of testing of AY-102. Lindberg and Deutsch (2003) described the first phase of testing on AY-102 material

Structural analysis of multiport riser 5A installation on tank 241SY101

Energy Technology Data Exchange (ETDEWEB)

Strehlow, J.P.

1994-09-16

The Tank 101-SY multiport riser assembly in the 241-SY-101 waste tank will replace the existing 42 inch riser with four smaller ports. Each smaller port can be used independently to access the tank interior with equipment and instruments needed to mitigate the concentration of hydrogen in the tank. This document provides a design report on the structural evaluation of the multiport riser assembly as well as its anchorage. The multiport riser assembly is a steel structure installed directly above the 42-inch riser and sealed at the existing riser flange. The assembly is structurally supported by the concrete pad placed around the 42 inch riser. The multiport riser assembly will provide two 8-inch penetrations, one 12-inch penetration and one 24-inch penetration. Each penetration will have a shielding plate. These penetrations will be used to insert equipment such as a sonic probe into the tank. In addition to normal loads, non-reactor Safety Class 1 structures, systems and components are to withstand the effects of extreme environmental loads including Design Basis Earthquake (DBE), Design Basis Wind (DBW), Design Basis Flood, Volcanic Eruptions and other abnormal loads considered on a case by case basis. Non-reactor Safety Class 2, 3 and 4 structures, systems and components are those that are not Safety Class 1 and are respectively specified as onsite safety related, occupational safety related and non-safety related items. The 241-SY-101 tank is considered as a non-reactor Safety Class 1 structure. The multiport riser assembly is considered as a non-reactor Safety Class 2 structure since it serves to contain the radioactive and toxic materials under normal operating conditions. However, the pressure relief doors provided on the assembly are considered as Safety Class 1 structures.

Laboratory testing of ozone oxidation of Hanford Site waste from Tank 241-SY-101

International Nuclear Information System (INIS)

Delegard, C.H.; Stubbs, A.M.; Bolling, S.D.

1993-01-01

Ozone was investigated as a reagent to oxidize and destroy organic species present in simulated and genuine waste from Hanford Site Tank 241-SY-101 (Tank 101-SY). Two high-shear mixing apparatus were tested to perform the gas-to-solution mass transfer necessary to achieve efficient use of the ozone reagent. Oxidations of nitrite (to form nitrate) and organic species were observed. The organics oxidized to form carbonate and oxalate as well as nitrate and nitrogen gas from nitrogen associated with the organic. oxidations of metal species also were observed directly or inferred by solubilities. The chemical reaction stoichiometries were consistent with reduction of one oxygen atom per ozone molecule. Acetate, oxalate, and formate were found to comprise about 40% of the genuine waste's total organic carbon (TOC) concentration. Ozonation was found to be chemically feasible for destroying organic species (except oxalate) present in the wastes in Tank 101-SY. The simulated waste formulation used in these studies credibly modelled the ozonation behavior of the genuine waste

Operational test report, integrated system test (ventilation upgrade)

International Nuclear Information System (INIS)

HARTY, W.M.

1999-01-01

Operational Final Test Report for Integrated Systems, Project W-030 (Phase 2 test, RECIRC and HIGH-HEAT Modes). Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks, including upgraded vapor space cooling and filtered venting of tanks AY101, AY102, AZ101, AZ102

Operational test report integrated system test (ventilation upgrade)

Energy Technology Data Exchange (ETDEWEB)

HARTY, W.M.

1999-10-05

Operational Final Test Report for Integrated Systems, Project W-030 (Phase 2 test, RECIRC and HIGH-HEAT Modes). Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks, including upgraded vapor space cooling and filtered venting of tanks AY101, Ay102, AZ101, AZ102.

Ventilation system consequence calculations to support salt well pumping single-shell tank 241-A-101

Energy Technology Data Exchange (ETDEWEB)

Ryan, G.W.

1997-05-07

This document presents the radiological dose and toxicological exposure calculations for an accident scenario involved with the ventilation system used to support salt well pumping single-shell tank 241-A-101. This tank has been listed on the Hydrogen Watch List.

Ventilation system consequence calculations to support salt well pumping single-shell tank 241-A-101

International Nuclear Information System (INIS)

Ryan, G.W.

1997-01-01

This document presents the radiological dose and toxicological exposure calculations for an accident scenario involved with the ventilation system used to support salt well pumping single-shell tank 241-A-101. This tank has been listed on the Hydrogen Watch List

Ultrasonic Examination of Double-Shell Tank 241-AY-101 Examination Completed August 2003

International Nuclear Information System (INIS)

Pardini, Allan F.; Posakony, Gerald J.

2003-01-01

COGEMA Engineering Corporation (COGEMA), under a contract from CH2M Hill Hanford Group (CH2M Hill), has performed an ultrasonic nondestructive examination of selected portions of Double-Shell Tank 241-AY-101. The purpose of this examination was to provide information that could be used to evaluate the integrity of the wall of the secondary tank. The requirements for the ultrasonic examination of Tank 241-AY-101 were to detect, characterize (identify, size, and locate), and record measurements made of any wall thinning or pitting that might be present in the wall of the secondary tank. Any measurements that exceed the requirements set forth in the Engineering Task Plan (ETP), RPP--11832 (Jensen 2002) and summarized on page 1 of this document, are reported to CH2M Hill and the Pacific Northwest National Laboratory (PNNL) for further evaluation. Under the contract with CH2M Hill, all data is to be recorded on disk and paper copies of all measurements are provided to PNNL for third-party evaluation. PNNL is responsible for preparing a report that describes the results of the COGEMA ultrasonic examinations

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

Soil structure interaction analysis for the Hanford Site 241-SY-101 double-shell waste storage tanks

International Nuclear Information System (INIS)

Giller, R.A.; Weiner, E.O.

1991-09-01

The 241-SY-101 tank is a double-shell waste storage tank buried in the 241-SY tank farm in the 200 West Area of the Hanford Site. This analysis addresses the effects of seismic soil-structure interaction on the tank structure and includes a parametric soil-structure interaction study addressing three configurations: two-dimensional soil structure, a two-dimensional structure-soil-structure, and a three-dimensional soil-structure interaction. This study was designed to determine an optimal method for addressing seismic-soil effects on underground storage tanks. The computer programs calculate seismic-soil pressures on the double-shell tank walls and and seismic acceleration response spectra in the tank. The results of this soil-structure interaction parametric study as produced by the computer programs are given in terms of seismic soil pressures and response spectra. The conclusions of this soil-structure interaction evaluation are that dynamically calculated soil pressures in the 241-SY-101 tank are significantly reduce from those using standard hand calculation methods and that seismic evaluation of underground double-shell waste storage tanks must consider soil-structure interaction effects in order to predict conservative structural response. Appendixes supporting this study are available in Volume 2 of this report

Vapor space characterization of waste Tank 241-TY-101: Results from samples collected on 4/6/95

International Nuclear Information System (INIS)

Klinger, G.S.; Clauss, T.W.; Ligotke, M.W.; Pool, K.H.; McVeety, B.D.; Olsen, K.B.; Bredt, O.P.; Fruchter, J.S.; Goheen, S.C.

1995-11-01

This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-TY-101 (referred to as Tank TY-101). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water vapor (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Off these, 5 were observed above the 5-ppbv reporting cutoff. One tentatively identified compound (TIC) was observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The six organic analyses identified are listed in Table 1 and account for approximately 100% of the total organic components in Tank TY-101. Two permanent gases, carbon dioxide (CO 2 ) and nitrous oxide (N 2 O), were also detected. Tank TY-101 is on the Ferrocyanide Watch List

Process Control Plan for Tank 241-SY-101 Surface Level Rise Remediation

International Nuclear Information System (INIS)

ESTEY, S.D.

1999-01-01

The tank 241-SY-101 transfer system was conceived and designed to address the immediate needs presented by rapidly changing waste conditions in tank 241-SY-101. Within approximately the last year, the waste in this tank has exhibited unexpected behavior (Rassat et al. 1999) in the form of rapidly increasing crust growth. This growth has been brought about by a rapidly increasing rate of gas entrapment within the crust. It has been conceived that the lack of crust agitation beginning upon the advent of mixer pump operations may have set-up a more consolidated, gas impermeable barrier when compared to a crust regularly broken up by the prior buoyant displacement events within the tank. The interim goals of the project are to: (1) protect the mixer pump operability (2) begin releasing gas from the crust, and (3) begin dissolving the crust and solids in the slurry layer. The final goals of the project (Final State) are to solve both the level growth and BD-GRE safety issues in this tank by achieving a condition of the waste such that no active measures are required to safely store the waste, i.e., crust and non convective layer are mostly dissolved, and therefore the mixer pump will no longer be needed to prevent BD-GREs in excess of 100% LFL. Transfers (which are designed to create space in the tank) and dilution (which will dissolve the solids) will accomplish this. Dissolution of solids will result in a release of gas retained by those solids and remove that volume of solids as a future retention site

Final report for Tank 241-B-101, push mode cores 90 and 91. Revision 1

International Nuclear Information System (INIS)

Schreiber, R.D.

1995-11-01

This is the final report for tank 241-at sign 101, cores 90 and 91. Samples from these cores were analyzed for safety screening purposes in accordance with the Tank 241-B-101 Tank Characterization Plan (TCP) (Reference 1). This final report contains the results for three sets of TGA and gravimetric analyses performed after the 90-day report was issued. Two of these TGA/gravimetric percent water sets of analyses were done because low original TGA results were obtained for the lower half segment of core 90, segment 2 and the facie of core 91, segment 2; the third set of analyses were performed because the TGA and gravimetric percent water results for the upper half segment of core 90, segment 2 differed by approximately a factor of three and further investigation was warranted

Miscellaneous component design for Tank 241SY101 pump removal

International Nuclear Information System (INIS)

Huang, F.H.

1995-01-01

A mixer pump has been used to mitigate the hydrogen build-up in tank 241SY101 (SY101), located in the 200 West Area of the Hanford Site. New equipment is being prepared for the removal, transport, storage, and disposal of the test pump. The disposal equipment for the test pump now in tank SY101 includes a shipping container, a strong back, a lifting beam, a test weight, container support stands, a modified mock-up pump, a flexible receiver blast shield, a lifting yoke, and a yoke brace. The structural evaluations of container and strong back are detailed in another supporting document (WHC 1994a), the engineering analyses of flexible receiver blast shield/lifting yoke and yoke brace are given in other supporting documents (WHC 1994b, WHC 1994c), respectively. Engineering tasks that were contracted to Advanced Engineering Consultants (AEC) include the design and analysis of the following. Two spreader-beam lifting devices. a Container test weight. Container support saddles. Mock-up pump modification. This report documents the work description, design basis, assumptions, and design calculations provided by AEC for the above components. All AEC documents appear in Appendix A. Additional work conducted by Westinghouse Hanford Company (WHC) on the modified container test weight, modification to the mock-up pump, the removable support for the transport assembly, and saddle modification for air pallets also are included in this document

Thermal hydraulic behavior evaluation of tank A-101

International Nuclear Information System (INIS)

Ogden, D.M.

1996-01-01

This report describes a new evaluation conducted to help understand the thermal-hydraulic behavior of tank A-101. Prior analysis of temperature data indicated that the dome space and upper waste layer was slowly increasing in temperature increases are due to increasing ambient temperatures and termination of forced ventilation. However, this analysis also indicates that other dome cooling processes are slowly decreasing, or some slow increase in heating is occurring at the waste surface. Dome temperatures are not decreasing at the rate expected as a forced ventilation termination effects are accounted for

Preoperational test report, primary ventilation system

Energy Technology Data Exchange (ETDEWEB)

Clifton, F.T.

1997-11-04

This represents a preoperational test report for Primary Ventilation Systems, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system provides vapor space filtered venting of tanks AY101, AY102, AZ101, AZ102. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System.

Preoperational test report, primary ventilation system

International Nuclear Information System (INIS)

Clifton, F.T.

1997-01-01

This represents a preoperational test report for Primary Ventilation Systems, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system provides vapor space filtered venting of tanks AY101, AY102, AZ101, AZ102. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System

Operability test procedure [Tank] 241-SY-101 equipment removal system

International Nuclear Information System (INIS)

Mast, J.C.

1994-01-01

The 241-SY-101 equipment removal system (ERS) consists of components, equipment, instrumentation and procedures that will provide the means to disconnect, retrieve, contain, load and transport the Mitigation Pump Assembly (MPA) from waste Tank 241-SY-101 to the Central Waste Complex (CWC). The Operability Test Procedure (OTP) will test the interfaces between ERS components and will rehearse the procedure for MPA removal and transportation to the extent they can be mocked-up at the CTF (Cold Test Facility). At the conclusion of the OTP, the ERS components and equipment will be removed from the CTF, entered into the Component Based Recall System (CBRS), and stored until needed for actual MPA removal and transportation

Tank 101-SY Window E core sample: Interpretation of results

International Nuclear Information System (INIS)

Reynolds, D.A.

1993-02-01

A full depth core sample was taken for tank 241-SY-101 in December 1991 during a time period called ''Window E.'' This was the second full depth core sample from this tank during the year. The core had two major portions that are known as the convective zone and the nonconvective zone. A crust was on the top of tank but as poorly sampled. The analysis of the Window E core sample stressed segment composite chemical analysis instead of segment by segment as in Window C. Adiabatic calorimetry on samples from both cores showed a slow self heating reaction above 150 degrees C on dried samples. The exothermic events were milder than similar synthetic samples. The chemical and physical properties complemented the information from Window C. The Window E material from the convective zone was more viscous than the Window C convective zone material. The nonconvective zone viscosities were similar for both cores. Heating and dilution tests were made to test mitigation concepts

Tank 101-SY Window E core sample: Interpretation of results

Energy Technology Data Exchange (ETDEWEB)

Reynolds, D.A.

1993-02-01

A full depth core sample was taken for tank 241-SY-101 in December 1991 during a time period called ``Window E.`` This was the second full depth core sample from this tank during the year. The core had two major portions that are known as the convective zone and the nonconvective zone. A crust was on the top of tank but as poorly sampled. The analysis of the Window E core sample stressed segment composite chemical analysis instead of segment by segment as in Window C. Adiabatic calorimetry on samples from both cores showed a slow self heating reaction above 150{degrees}C on dried samples. The exothermic events were milder than similar synthetic samples. The chemical and physical properties complemented the information from Window C. The Window E material from the convective zone was more viscous than the Window C convective zone material. The nonconvective zone viscosities were similar for both cores. Heating and dilution tests were made to test mitigation concepts.

Tank Vapor Characterization Project: Headspace vapor characterization of Hanford waste tank 241-S-101: Results from samples collected on 06/06/96

International Nuclear Information System (INIS)

Thomas, B.L.; Evans, J.C.; Pool, K.H.; Olsen, K.B.; Fruchter, J.S.; Silvers, K.L.

1997-01-01

This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-S-101. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained. Analyte concentrations were based on analytical results and sample volumes provided by WHC. A summary of the inorganic analytes, permanent gases, and total non-methane organic compounds is listed

241-AZ-101 Mixer Pump Demonstration Test Gamma Cart Acceptance Test Procedure and Quality Test Plan (ATP and QTP)

International Nuclear Information System (INIS)

WHITE, D.A.

2000-01-01

Shop test of the sludge mobilization cart system to be used in the AZ-101 Mixer Pump Demonstration Test Tests hardware and software. This procedure involves testing the Instrumentation involved with the Gamma Cart System, local and remote, including depth indicators, speed controls, interface to data acquisition software and the raising and lowering functions. This Procedure will be performed twice, once for each Gamma Cart System. This procedure does not test the accuracy of the data acquisition software

241-AZ-101 Mixer Pump Demonstration Test Gamma Cart Acceptance Test Procedure and Quality Test Plan (ATP and QTP)

International Nuclear Information System (INIS)

WHITE, D.A.

2000-01-01

Shop Test of the Gamma Cart System to be used in the AZ-101 Mixer Pump Demonstration Test. Tests hardware and software. This procedure involves testing the Instrumentation involved with the Gamma Cart System, local and remote, including: depth indicators, speed controls, interface to data acquisition software and the raising and lowering functions. This Procedure will be performed twice, once for each Gamma Cart System. This procedure does not test the accuracy of the data acquisition software

Integrity assessment report of tanks TK-101 and TK-102

Energy Technology Data Exchange (ETDEWEB)

MCSHANE, D.S.

1999-08-25

This Integrity Assessment Report (IAR) is prepared by Fluor Daniel Northwest (FDNW) for Waste Management Federal Services of Hanford, Inc., (WMH), the operations contractor; Fluor Daniel Hanford (FDH), the Hanford Site Manager; and the U. S. Department of Energy (DOE), the system owner. This IAR addresses the evaluation of Tanks 101 and 102 and other existing components located in the 219-S Waste Handling Facility. This report will be included in the Part B Permit for the 2226 Laboratory and is a portion of the integrity assessment of the overall 222-5 Laboratory radioactive liquid waste disposal system. This IAR is prepared in accordance with WAC 173-303, Dangerous Waste Regulations; Section 640(2), ''Assessment of Existing Tank Systems Integrity .''

Integrity assessment report of tanks TK-101 and TK-102

International Nuclear Information System (INIS)

MCSHANE, D.S.

1999-01-01

This Integrity Assessment Report (IAR) is prepared by Fluor Daniel Northwest (FDNW) for Waste Management Federal Services of Hanford, Inc., (WMH), the operations contractor; Fluor Daniel Hanford (FDH), the Hanford Site Manager; and the U. S. Department of Energy (DOE), the system owner. This IAR addresses the evaluation of Tanks 101 and 102 and other existing components located in the 219-S Waste Handling Facility. This report will be included in the Part B Permit for the 2226 Laboratory and is a portion of the integrity assessment of the overall 222-5 Laboratory radioactive liquid waste disposal system. This IAR is prepared in accordance with WAC 173-303, Dangerous Waste Regulations; Section 640(2), ''Assessment of Existing Tank Systems Integrity .''

Ultrasonic Examination of Double-Shell Tank 241-AY-101. Examination completed October 2007

International Nuclear Information System (INIS)

Pardini, Allan F.; Weier, Dennis R.

2008-01-01

AREVA NC Inc., under contract from CH2M Hill Hanford Group, has performed an ultrasonic examination of selected portions of Double-Shell Tank 241-AY-101. PNNL is responsible for preparing a report(s) that describes the results of the AREVA ultrasonic examinations. This report is Revision 1 - more data has been added to the original report. The purpose of this examination was to provide information that could be used to evaluate the integrity of the wall of the primary tank. The requirements for the ultrasonic examination of Tank 241-AY-101 were to detect, characterize (identify, size, and locate), and record measurements made of any wall thinning, pitting, or cracks that might be present in the wall of the primary tank. Any measurements that exceed the requirements set forth in the Engineering Task Plan, RPP-Plan-27202 (Jensen 2005) and summarized on page 1 of this document, are to be reported to CH2M Hill Hanford Group and the Pacific Northwest National Laboratory for further evaluation. Under the contract with CH2M Hill Hanford Group, all data is to be recorded on electronic media and paper copies of all measurements are provided to Pacific Northwest National Laboratory for third-party evaluation. Pacific Northwest National Laboratory is responsible for preparing a report(s) that describes the results of the AREVA NC Inc. ultrasonic examinations.

Similarity analysis applied to the design of scaled tests of hydraulic mitigation methods for Tank 241-SY-101

International Nuclear Information System (INIS)

Liljegren, L.M.

1993-02-01

The episodic gas releases from Tank 241-SY-101 (SY-101) pose a potential safety hazard. It is thought that gas releases occur because gases are generated and trapped in layers of settled solids located at the bottom of the tank. This document focuses on issues associated with testing of hydraulic mitigation technologies proposed for SY-101. The basic assumption underlying the concept of hydraulic mitigation is that mobilization or maintained suspension of the solids settled in the bottom of the tank wig prevent gas accumulation. Engineering of hydraulic technologies will require testing to determine the operating parameters required to mobilize the solids and to maintain these solids in suspension. Because full scale testing is extremely expensive (even when possible), scaled tests are needed to assess the merit of the proposed technologies and to provide data for numerical or analytical modeling. This research is conducted to support testing and evaluation of proposed hydraulic mitigation concepts only. The work here is oriented towards determining the jet velocities, nozzle sizes, and other operating parameters required to mobilize the settled solids in SY- 101 and maintain them in suspension

Assessment of the potential for ammonium nitrate formation and reaction in Tank 241-SY-101

International Nuclear Information System (INIS)

Pederson, L.R.; Bryan, S.A.

1994-08-01

Two principal scenarios by which ammonium nitrate may be formed were considered: (a) precipitation of ammonium nitrate in the waste, and (b) ammonium nitrate formation via the gas phase reaction of ammonia and nitrogen dioxide. The first of these can be dismissed because ammonium ions, which are necessary for ammonium nitrate precipitation, can exist only in negligibly small concentrations in strongly alkaline solutions. Gas phase reactions between ammonia, nitrogen dioxide, and water vapor in the gas phase represent the most likely means by which ammonium nitrate aerosols could be formed in Tank 241-SY-101. Predicted ammonium nitrate formation rates are largely controlled by the concentration of nitrogen dioxide. This gas has not been detected among those gases vented from the wastes using Fourier Transform Infrared Spectrometry (FTIR) or mass spectrometry. While detection limits for nitrogen dioxide have not been established experimentally, the maximum concentration of nitrogen dioxide in the gas phase in Tank 241-SY-101 was estimated at 0.1 ppm based on calculations using the HITRAN data base and on FTIR spectra of gases vented from the wastes. At 50 C and with 100 ppm ammonia also present, less than one gram of ammonium nitrate per year is estimated to be formed in the tank. To date, ammonium nitrate has not been detected on HEPA filters in the ventilation system, so any quantity that has been formed in the tank must be quite small, in good agreement with rate calculations. The potential for runaway exothermic reactions involving ammonium nitrate in Tank 241-SY-101 is minimal. Dilution by non-reacting waste components, particularly water, would prevent hazardous exothermic reactions from occurring within the waste slurry, even if ammonium nitrate were present. 41 refs

Test Plan And Procedure For The Examination Of Tank 241-AY-101 Multi-Probe Corrosion Monitoring System

International Nuclear Information System (INIS)

Wyrwas, R.B.; Page, J.S.; Cooke, G.S.

2012-01-01

This test plan describes the methods to be used in the forensic examination of the Multi-probe Corrosion Monitoring System (MPCMS) installed in the double-shell tank 241-AY-101 (AY-101). The probe was designed by Applied Research and Engineering Sciences (ARES) Corporation. The probe contains four sections, each of which can be removed from the tank independently (H-14-107634, AY-101 MPCMS Removable Probe Assembly) and one fixed center assembly. Each removable section contains three types of passive corrosion coupons: bar coupons, round coupons, and stressed C-rings (H-14-l07635, AY-101 MPCMS Details). Photographs and weights of each coupon were recorded and reported on drawing H-14-107634 and in RPP-RPT-40629, 241-AY-101 MPCMS C-Ring Coupon Photographs. The coupons will be the subject of the forensic analyses. The purpose of this examination will be to document the nature and extent of corrosion of the 29 coupons. This documentation will consist of photographs and photomicrographs of the C-rings and round coupons, as well as the weights of the bar and round coupons during corrosion removal. The total weight loss of the cleaned coupons will be used in conjunction with the surface area of each to calculate corrosion rates in mils per year. The bar coupons were presumably placed to investigate the liquid-air-interface. An analysis of the waste level heights in the waste tank will be investigated as part of this examination.

A discussion on the safety classification of the tank 241-SY-101 mixer pump

International Nuclear Information System (INIS)

Van Vleet, R.J.

1997-01-01

An analysis, consistent with the methodology used in the draft TWRS FSAR (HNF-SD-WM-SAR-067), is presented to show that the classification of the mixer pump in tank 241-SY-101 should be safety significant

Tank 241-AY-101 Privatization Push Mode Core Sampling and Analysis Plan

International Nuclear Information System (INIS)

TEMPLETON, A.M.

2000-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AY-101. The purpose of this sampling event is to obtain information about the characteristics of the contents of 241-AY-101 required to satisfy Data Quality Objectives For RPP Privatization Phase I: Confirm Tank T Is An Appropriate Feed Source For High-Level Waste Feed Batch X(HLW DQO) (Nguyen 1999a), Data Quality Objectives For TWRS Privatization Phase I : Confirm Tank T Is An Appropriate Feed Source For Low-Activity Waste Feed Batch X (LAW DQO) (Nguyen 1999b), Low Activity Waste and High-Level Waste Feed Data Quality Objectives (L and H DQO) (Patello et al. 1999), and Characterization Data Needs for Development, Design, and Operation of Retrieval Equipment Developed through the Data Quality Objective Process (Equipment DQO) (Bloom 1996). Special instructions regarding support to the LAW and HLW DQOs are provided by Baldwin (1999). Push mode core samples will be obtained from risers 15G and 150 to provide sufficient material for the chemical analyses and tests required to satisfy these data quality objectives. The 222-S Laboratory will extrude core samples; composite the liquids and solids; perform chemical analyses on composite and segment samples; archive half-segment samples; and provide subsamples to the Process Chemistry Laboratory. The Process Chemistry Laboratory will prepare test plans and perform process tests to evaluate the behavior of the 241-AY-101 waste undergoing the retrieval and treatment scenarios defined in the applicable DQOs. Requirements for analyses of samples originating in the process tests will be documented in the corresponding test plans and are not within the scope of this SAP

Tank 241-AY-101 Privatization Push Mode Core Sampling and Analysis Plan

International Nuclear Information System (INIS)

TEMPLETON, A.M.

2000-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AY-101. The purpose of this sampling event is to obtain information about the characteristics of the contents of 241-AY-101 required to satisfy ''Data Quality Objectives For RPP Privatization Phase I: Confirm Tank T Is An Appropriate Feed Source For High-Level Waste Feed Batch X(HLW DQO)' (Nguyen 1999a), ''Data Quality Objectives For TWRS Privatization Phase I: Confirm Tank T Is An Appropriate Feed Source For Low-Activity Waste Feed Butch X (LAW DQO) (Nguyen 1999b)'', ''Low Activity Waste and High-Level Waste Feed Data Quality Objectives (L and H DQO)'' (Patello et al. 1999), and ''Characterization Data Needs for Development, Design, and Operation of Retrieval Equipment Developed through the Data Quality Objective Process (Equipment DQO)'' (Bloom 1996). Special instructions regarding support to the LAW and HLW DQOs are provided by Baldwin (1999). Push mode core samples will be obtained from risers 15G and 150 to provide sufficient material for the chemical analyses and tests required to satisfy these data quality objectives. The 222-S Laboratory will extrude core samples; composite the liquids and solids; perform chemical analyses on composite and segment samples; archive half-segment samples; and provide sub-samples to the Process Chemistry Laboratory. The Process Chemistry Laboratory will prepare test plans and perform process tests to evaluate the behavior of the 241-AY-101 waste undergoing the retrieval and treatment scenarios defined in the applicable DQOs. Requirements for analyses of samples originating in the process tests will be documented in the corresponding test plans and are not within the scope of this SAP

Gas generation and retention in Tank 101-SY: A summary of laboratory studies, tank data, and information needs

International Nuclear Information System (INIS)

Pederson, L.R.; Ashby, E.C.; Jonah, C.; Meisel, D.; Strachan, D.M.

1992-06-01

Chemical and radioactive wastes from processes used to separate plutonium from uranium are stored in underground tanks at the Hanford Site in Washington state. In March 1981, it was observed that the volume of wastes in Tank 101-SY slowly increased, followed by a rapid decrease and the venting of large quantities of gases. These cycles occurred every 8 to 15 weeks and continue to the present time. Subsequent analyses showed that these gases were composed primarily of hydrogen and nitrous oxide (N 2 O). In response to the potential for explosion and release of hazardous materials to the environment, laboratory programs were initiated at Argonne National Laboratory (ANL), Georgia Institute of Technology (GIT), Pacific Northwest Laboratory (PNL), and Westinghouse Hanford Company (WHC), to develop a better understanding of the physical and chemical processes occurring in this waste tank. An aggressive sampling and analysis effort is also under way to characterize the wastes as fully as possible. These efforts will provide a technically defensible basis for safety analyses and future mitigation/remediation of the tank and its contents

Gas generation and retention in Tank 101-SY: A summary of laboratory studies, tank data, and information needs

Energy Technology Data Exchange (ETDEWEB)

Pederson, L.R. [comp.] [Pacific Northwest Lab., Richland, WA (United States); Ashby, E.C. [Georgia Inst. of Tech., Atlanta, GA (United States); Jonah, C.; Meisel, D. [Argonne National Lab., IL (United States); Strachan, D.M. [Pacific Northwest Lab., Richland, WA (United States)

1992-06-01

Chemical and radioactive wastes from processes used to separate plutonium from uranium are stored in underground tanks at the Hanford Site in Washington state. In March 1981, it was observed that the volume of wastes in Tank 101-SY slowly increased, followed by a rapid decrease and the venting of large quantities of gases. These cycles occurred every 8 to 15 weeks and continue to the present time. Subsequent analyses showed that these gases were composed primarily of hydrogen and nitrous oxide (N{sub 2}O). In response to the potential for explosion and release of hazardous materials to the environment, laboratory programs were initiated at Argonne National Laboratory (ANL), Georgia Institute of Technology (GIT), Pacific Northwest Laboratory (PNL), and Westinghouse Hanford Company (WHC), to develop a better understanding of the physical and chemical processes occurring in this waste tank. An aggressive sampling and analysis effort is also under way to characterize the wastes as fully as possible. These efforts will provide a technically defensible basis for safety analyses and future mitigation/remediation of the tank and its contents.

Composition, preparation, and gas generation results from simulated wastes of Tank 241-SY-101

International Nuclear Information System (INIS)

Bryan, S.A.; Pederson, L.R.

1994-08-01

This document reviews the preparation and composition of simulants that have been developed to mimic the wastes temporarily stored in Tank 241-SY-101 at Hanford. The kinetics and stoichiometry of gases that are generated using these simulants are also compared, considering the roles of hydroxide, chloride, and transition metal ions; the identities of organic constituents; and the effects of dilution, radiation, and temperature. Work described in this report was conducted for the Flammable Gas Safety Program at Pacific Northwest Laboratory, (a) whose purpose is to develop information that is necessary to mitigate potential safety hazards associated with waste tanks at the Hanford Site. The goal of this research and of related efforts at the Georgia Institute of Technology (GIT), Argonne National Laboratory (ANL), and Westinghouse Hanford Company (WHC) is to determine the thermal and thermal/radiolytic mechanisms by which flammable and other gases are produced in Hanford wastes, emphasizing those stored in Tank 241-SY-101. A variety of Tank 241-SY-101 simulants have been developed to date. The use of simulants in laboratory testing activities provides a number of advantages, including elimination of radiological risks to researchers, lower costs associated with experimentation, and the ability to systematically alter simulant compositions to study the chemical mechanisms of reactions responsible for gas generation. The earliest simulants contained the principal inorganic components of the actual waste and generally a single complexant such as N-(2-hydroxyethyl) ethylenediaminetriacetic acid (HEDTA) or ethylenediaminetriacetic acid (EDTA). Both homogeneous and heterogeneous compositional forms were developed. Aggressive core sampling and analysis activities conducted during Windows C and E provided information that was used to design new simulants that more accurately reflected major and minor inorganic components

702AZ aging waste ventilation facility year 2000 test procedure

International Nuclear Information System (INIS)

Winkelman, W.D.

1998-01-01

This test procedure was developed to determine if the 702AZ Tank Ventilation Facility system is Year 2000 Compliant. The procedure provides detailed instructions for performing the operations necessary and documenting the results. This verification procedure will document that the 702AZ Facility Systems are year 2000 compliant and will correctly meet the criteria established in this procedure

HLW Feed Delivery AZ101 Batch Transfer to the Private Contractor Transfer and Mixing Process Improvements [Initial Release at Rev 2

Energy Technology Data Exchange (ETDEWEB)

DUNCAN, G.P.

2000-02-28

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.

Progress toward mitigation of flammable gas Tank 241-SY-101

International Nuclear Information System (INIS)

Lentsch, J.W.; Babad, H.; Hanson, C.E.; Kirch, N.W.

1994-01-01

The mixing pump installed in Hanford Site tank 241-SY-101 has been shown to be effective in releasing flammable gases in a controlled manner. This controlled release of gas prevents the accumulation and episodic release above flammable limits. More work needs to be done to optimize the pumping operation, and to evaluate the long-term effects of mixing so as to assure that no undesirable changes have occurred to the waste. Other alternative mitigation concepts are still being evaluated as a backup to mixing

Korozní odolnost tvářených hořčíkových slitin AZ31 a AZ61 v Hankovì roztoku

Czech Academy of Sciences Publication Activity Database

Tkacz, J.; Slouková, K.; Minda, J.; Drábiková, J.; Fintová, Stanislava; Doležal, P.; Wasserbauer, J.

2016-01-01

Roč. 60, č. 4 (2016), s. 101-106 ISSN 1804-1213 Institutional support: RVO:68081723 Keywords : corrosion * AZ31 magnesium alloy * AZ61 magnesium alloy Subject RIV: JK - Corrosion ; Surface Treatment of Materials https://www.degruyter.com/view/j/kom.2016.60.issue-4/kom-2016-0016/kom-2016-0016.xml?format=INT

Solubilities of gases in simulated Tank 241-SY-101 wastes

International Nuclear Information System (INIS)

Norton, J.D.; Pederson, L.R.

1995-09-01

Oxygen, nitrogen, hydrogen, methane, and nitrous oxide solubilities were evaluated as a function of temperature in SYl-SIM-93B, a homogeneous simulated waste mixture containing sodium hydroxide, sodium nitrite, sodium nitrate, sodium aluminate, and sodium carbonate, the principal inorganic constituents of the wastes in Tank 241-SY-101. Ammonia solubility data for this simulated waste was obtained as a function of temperature in an earlier study. The choice of a homogeneous waste mixture in this study has the advantage of eliminating complications associated with a changing electrolyte concentration as a function of temperature that would be encountered with a slurry simulant. Dissolution is one of the means by which gases may be retained in Hanford Site wastes. While models are available to estimate gas solubilities in electrolyte solutions, few data are in existence that pertain to highly concentrated, multicomponent electrolytes such as those stored in Hanford Site waste tanks

The behavior, quantity, and location of undissolved gas in Tank 241-SY-101

Energy Technology Data Exchange (ETDEWEB)

Brewster, M.E.; Gallagher, N.B.; Hudson, J.D.; Stewart, C.W.

1995-10-01

Mitigation of episodic flammable gas releases from Hanford Waste Tank 241-SY-101 was accomplished in July 1993 with the installation of a mixer pump that prevents gas retention. But is has not been possible until recently to measure the effects of mixing on the waste or how much gas remains and where it is located. Direct measurements of the void fraction and rheology of the mixed waste by the void fraction instrument (VFI) and ball rheometer along with previous data provide estimates of the location, quantity, and behavior of undissolved gas in the tank. This report documents the compilation and integration of the information that enables this understanding.

Evaluation of cracking in the 241-AZ tank farm ventilation line at the Hanford Site

Energy Technology Data Exchange (ETDEWEB)

ANANTATMULA, R.P.

1999-10-20

In the period from April to October of 1988, a series of welding operations on the outside of the AZ Tank Farm ventilation line piping at the Hanford Site produced unexpected and repeated cracking of the austenitic stainless steel base metal and of a seam weld in the pipe. The ventilation line is fabricated from type 304L stainless steel pipe of 24 inch diameter and 0.25 inch wall thickness. The pipe was wrapped in polyethylene bubble wrap and buried approximately 12 feet below grade. Except for the time period between 1980 and 1987, impressed current cathodic protection has been applied to the pipe since its installation in 1974. The paper describes the history of the cracking of the pipe, the probable cracking mechanisms, and the recommended future action for repair/replacement of the pipe.

ATR/OTR-SY Tank Camera Purge System and in Tank Color Video Imaging System

International Nuclear Information System (INIS)

Werry, S.M.

1995-01-01

This procedure will document the satisfactory operation of the 101-SY tank Camera Purge System (CPS) and 101-SY in tank Color Camera Video Imaging System (CCVIS). Included in the CPRS is the nitrogen purging system safety interlock which shuts down all the color video imaging system electronics within the 101-SY tank vapor space during loss of nitrogen purge pressure

Evaluation of the generation and release of flammable gases in tank 241-SY-101

Energy Technology Data Exchange (ETDEWEB)

Babad, H.; Johnson, G.D.; Lechelt, J.A.; Reynolds, D.A. (Westinghouse Hanford Co., Richland, WA (United States)); Pederson, L.R.; Strachan, D.M. (Pacific Northwest Lab., Richland, WA (United States)); Meisel, D.; Jonah, C. (Argonne National Lab., IL (United States)); Ashby, E.C. (Georgia Inst. of Tech., Atlanta, GA (United States))

1991-11-01

Tank 241-SY-101 is a double shell, high-level waste tank located in the 200 West Area of the Hanford Site. This tank contains about 1 million gallons of waste that was concentrated at the 242-S Evaporator. Shortly after the waste was put in the tank, the waste began to expand because the generation of gases. In 1990 this tank was declared to have an unreviewed safety question because of the periodic release of hydrogen and nitrous oxide. A safety program was established to conduct a characterization of the waste and vented gases and to determine an effective means to prevent the accumulation of flammable gases in the tank dome space and ventilation system. Results of the expanded characterization conducted in fiscal year 1991 are presented. The use of gas chromatographs, mass spectrometers, and hydrogen-specific monitors provided a greater understanding of the vented gases. Additional instrumentation placed in the tank also helped to provide more detailed information on tank temperatures, gas pressure, and gas flow rates. An extensive laboratory study involving the Westinghouse Hanford Company, Pacific Northwest Laboratory, Argonne National Laboratory, and the Georgia Institute of Technology was initiated for the purpose of determining the mechanisms responsible for the generation of various gases. These studies evaluate both radiolytic and thermochemical processes. Results of the first series of experiments are described.

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)

45-Day safety screen results for Tank 241-C-101, auger sample 95-AUG-019

International Nuclear Information System (INIS)

Sasaki, L.M.

1995-01-01

One auger sample from Tank 241-C-101 was received by the 222-S Laboratory and underwent safety screening analyses--differential scanning calorimetry (DSC), thermogravimetric analysis (TGA), and total alpha analysis--in accordance with the tank characterization plan. Analytical results for the TGA on the crust sample (the uppermost portion of the auger sample) (sample number S95T000823) were less than the safety screening notification limit of 17 weight percent water. Verbal and written notifications were made on May 3, 1995. No exotherms were observed in the DSC analyses and the total alpha results were well below the safety screening notification limit. This report includes the primary safety screening results obtained from the analyses and copies of all DSC and TGA raw data scans as requested per the TCP. Although not included in this report, a photograph of the extruded sample was taken and is available. This report also includes bulk density measurements required by Characterization Plant Engineering. Additional analyses (pH, total organic carbon, and total inorganic carbon) are being performed on the drainable liquid at the request of Characterization Process Control; these analyses will be reported at a later date in a final report for this auger sample. Tank C-101 is not part of any of the four Watch Lists

Transport of Tank 241-SY-101 Waste Slurry: Effects of Dilution and Temperature on Critical Pipeline Velocity

International Nuclear Information System (INIS)

KP Recknagle; Y Onishi

1999-01-01

This report presents the methods and results of calculations performed to predict the critical velocity and pressure drop required for the two-inch pipeline transfer of solid/liquid waste slurry from underground waste storage Tank 241-SY-101 to Tank 241-SY- 102 at the Hanford Site. The effects of temperature and dilution on the critical velocity were included in the analysis. These analyses show that Tank 241-SY-101 slurry should be diluted with water prior to delivery to Tank 241-SY-102. A dilution ratio of 1:1 is desirable and would allow the waste to be delivered at a critical velocity of 1.5 ft/sec. The system will be operated at a flow velocity of 6 ft/sec or greater therefore, this velocity will be sufficient to maintain a stable slurry delivery through the pipeline. The effect of temperature on the critical velocity is not a limiting factor when the slurry is diluted 1:1 with water. Pressure drop at the critical velocity would be approximately two feet for a 125-ft pipeline (or 250-ft equivalent straight pipeline). At 6 ft/sec, the pressure drop would be 20 feet over a 250-ft equivalent straight pipeline

In situ determination of rheological properties and void fraction in Hanford Waste Tank 241-SY-101

International Nuclear Information System (INIS)

Stewart, C.W.; Shepard, C.L.; Alzheimer, J.M.; Stokes, T.I.; Terrones, G.

1995-08-01

This report presents the results of the operation of the void fraction instrument (VFI) and ball rheometer in Hanford Tank 241-SY-101, which contains approximately one million gallons of radioactive waste. These instruments provided the first direct assay of the waste condition in the tank after more than a year of mixer pump operation. The two instruments were deployed in the tank in late 1994 and early 1995 to gather much-needed data on the effect prolonged mixer pump operation has on gas retention in the waste. The information supplied by these instruments has filled a great gap in the quantitative knowledge of the waste condition. The results show that the solids are well-mixed by the current mixer pump to within less than a meter of the tank bottom. Undisturbed sludge remains only on the lowest 10--30 cm and contains 10--12% void. The mixed slurry above contains less than 1% void and has no measurable yield strength and a shear-thinning viscosity of approximately 6 Poise at 1 sec -1 . Estimating the gas volumes in each of the four layers based on VFI data yields a total of 221 ± 57 m 3 (7,800 ± 2,000 SCF) of gas at 1 atmosphere. Given the current waste level of 10.2 m (400 inches), the degassed waste level would be 9.8 m (386 inches). These results confirm that the mixer pump in Tank 241-SY-101 has performed the job it was installed to do--thoroughly mix the waste to release stored gas and prevent gas accumulation

Structural analysis of color video camera installation on tank 241AW101 (2 Volumes)

Energy Technology Data Exchange (ETDEWEB)

Strehlow, J.P.

1994-08-24

A video camera is planned to be installed on the radioactive storage tank 241AW101 at the DOE` s Hanford Site in Richland, Washington. The camera will occupy the 20 inch port of the Multiport Flange riser which is to be installed on riser 5B of the 241AW101 (3,5,10). The objective of the project reported herein was to perform a seismic analysis and evaluation of the structural components of the camera for a postulated Design Basis Earthquake (DBE) per the reference Structural Design Specification (SDS) document (6). The detail of supporting engineering calculations is documented in URS/Blume Calculation No. 66481-01-CA-03 (1).

Structural analysis of color video camera installation on tank 241AW101 (2 Volumes)

International Nuclear Information System (INIS)

Strehlow, J.P.

1994-01-01

A video camera is planned to be installed on the radioactive storage tank 241AW101 at the DOE' s Hanford Site in Richland, Washington. The camera will occupy the 20 inch port of the Multiport Flange riser which is to be installed on riser 5B of the 241AW101 (3,5,10). The objective of the project reported herein was to perform a seismic analysis and evaluation of the structural components of the camera for a postulated Design Basis Earthquake (DBE) per the reference Structural Design Specification (SDS) document (6). The detail of supporting engineering calculations is documented in URS/Blume Calculation No. 66481-01-CA-03 (1)

Evidence for dawsonite in Hanford high-level nuclear waste tanks.

Science.gov (United States)

Reynolds, Jacob G; Cooke, Gary A; Herting, Daniel L; Warrant, R Wade

2012-03-30

Gibbsite [Al(OH)(3)] and boehmite (AlOOH) have long been assumed to be the most prevalent aluminum-bearing minerals in Hanford high-level nuclear waste sludge. The present study shows that dawsonite [NaAl(OH)(2)CO(3)] is also a common aluminum-bearing phase in tanks containing high total inorganic carbon (TIC) concentrations and (relatively) low dissolved free hydroxide concentrations. Tank samples were probed for dawsonite by X-ray Diffraction (XRD), Scanning Electron Microscopy with Energy Dispersive Spectrometry (SEM-EDS) and Polarized Light Optical Microscopy. Dawsonite was conclusively identified in four of six tanks studied. In a fifth tank (AN-102), the dawsonite identification was less conclusive because it was only observed as a Na-Al bearing phase with SEM-EDS. Four of the five tank samples with dawsonite also had solid phase Na(2)CO(3) · H(2)O. The one tank without observable dawsonite (Tank C-103) had the lowest TIC content of any of the six tanks. The amount of TIC in Tank C-103 was insufficient to convert most of the aluminum to dawsonite (Al:TIC mol ratio of 20:1). The rest of the tank samples had much lower Al:TIC ratios (between 2:1 and 0.5:1) than Tank C-103. One tank (AZ-102) initially had dawsonite, but dawsonite was not observed in samples taken 15 months after NaOH was added to the tank surface. When NaOH was added to a laboratory sample of waste from Tank AZ-102, the ratio of aluminum to TIC in solution was consistent with the dissolution of dawsonite. The presence of dawsonite in these tanks is of significance because of the large amount of OH(-) consumed by dawsonite dissolution, an effect confirmed with AZ-102 samples. Copyright © 2012 Elsevier B.V. All rights reserved.

Mechanisms of gas retention and release: Experimental results for Hanford waste tanks 241-AW-101 and 241-AN-103

Energy Technology Data Exchange (ETDEWEB)

Rassat, S.D.; Gauglitz, P.A.; Bredt, P.R.; Mahoney, L.A.; Forbes, S.V.; Tingey, S.M.

1997-09-01

The 177 storage tanks at Hanford contain a vast array of radioactive waste forms resulting, primarily, from nuclear materials processing. Through radiolytic, thermal, and other decomposition reactions of waste components, gaseous species including hydrogen, ammonia, and the oxidizer nitrous oxide are generated within the waste tanks. Many of these tanks are known to retain and periodically release quantities of these flammable gas mixtures. The primary focus of the Flammable Gas Project is the safe storage of Hanford tank wastes. To this end, we strive to develop an understanding of the mechanisms of flammable gas retention and release in Hanford tanks through laboratory investigations on actual tank wastes. These results support the closure of the Flammable Gas Unreviewed Safety Question (USQ) on the safe storage of waste tanks known to retain flammable gases and support resolution of the broader Flammable Gas Safety Issue. The overall purpose of this ongoing study is to develop a comprehensive and thorough understanding of the mechanisms of flammable gas retention and release. The first objective of the current study was to classify bubble retention and release mechanisms in two previously untested waste materials from Tanks 241-AN-103 (AN-103) and 241-AW-101 (AW-101). Results were obtained for retention mechanisms, release characteristics, and the maximum gas retention. In addition, unique behavior was also documented and compared with previously studied waste samples. The second objective was to lengthen the duration of the experiments to evaluate the role of slowing bubble growth on the retention and release behavior. Results were obtained for experiments lasting from a few hours to a few days.

Mechanisms of gas retention and release: Experimental results for Hanford waste tanks 241-AW-101 and 241-AN-103

International Nuclear Information System (INIS)

Rassat, S.D.; Gauglitz, P.A.; Bredt, P.R.; Mahoney, L.A.; Forbes, S.V.; Tingey, S.M.

1997-09-01

The 177 storage tanks at Hanford contain a vast array of radioactive waste forms resulting, primarily, from nuclear materials processing. Through radiolytic, thermal, and other decomposition reactions of waste components, gaseous species including hydrogen, ammonia, and the oxidizer nitrous oxide are generated within the waste tanks. Many of these tanks are known to retain and periodically release quantities of these flammable gas mixtures. The primary focus of the Flammable Gas Project is the safe storage of Hanford tank wastes. To this end, we strive to develop an understanding of the mechanisms of flammable gas retention and release in Hanford tanks through laboratory investigations on actual tank wastes. These results support the closure of the Flammable Gas Unreviewed Safety Question (USQ) on the safe storage of waste tanks known to retain flammable gases and support resolution of the broader Flammable Gas Safety Issue. The overall purpose of this ongoing study is to develop a comprehensive and thorough understanding of the mechanisms of flammable gas retention and release. The first objective of the current study was to classify bubble retention and release mechanisms in two previously untested waste materials from Tanks 241-AN-103 (AN-103) and 241-AW-101 (AW-101). Results were obtained for retention mechanisms, release characteristics, and the maximum gas retention. In addition, unique behavior was also documented and compared with previously studied waste samples. The second objective was to lengthen the duration of the experiments to evaluate the role of slowing bubble growth on the retention and release behavior. Results were obtained for experiments lasting from a few hours to a few days

Nonradioactive Air Emissions Notice of Construction use of a portable exhauster on 241-A-101 tank during salt well pumping and other routine activities

International Nuclear Information System (INIS)

Hays, C.B.

1996-01-01

The 241-A-101 tank, a 22.9 meter 3,785,400 liter capacity SST, was constructed from the fourth generation of tank designs, which were capable of holding boiling or self-concentrating waste. Construction features a reinforced concrete shell, dome, and base with a mild steel liner covering the bottom and sidewalls. The tank has a flat bottom with a usable waste depth of approximately 9.4 meters. The tank was put into service in 1956 to store plutonium-uranium extraction (PUREX) high-level waste and organic wash waste. The waste was allowed to self-concentrate up until 1968. Tank sluicing was performed in 1969 and again in 1976 to reduce the amount of strontium and cesium, the two isotopes found to be the main heat generating sources in the tank. In 1978, the tank was reassigned for saltcake storage. The tank was taken out of service in November 1980 and partially isolated in 1982. Salt well pumping is a method used to interim stabilize SSTS. Interim stabilization is commenced once all the liquid above the solids has been removed (primary stabilization). Interim stabilization removes the gravity drainable liquid and the interstitial liquid between the solids from the SST and transfers the liquid to a double-shell tank (DST) or to a staging double-contained receiver tank (DCRT), which is subsequently transferred to a DST. Pumping is accomplished at very low flow rates, 15.1 liters per minute or less. Normally, salt well pumping is performed without the need of an exhauster. However, recent safety evaluations concluded that a minimum exhaust flow rate of 7.1 cubic meters per minute would be required to enhance the safety of the tank. Therefore, active ventilation will be part of this process for the 241-A-101 tank. This document details the Nonradioactive Air Emissions Notice of Construction for the use of a portable exhauster on Tank 241-A-101 during salt well pumping and other routine activities

Numerical simulation of Hanford Tank 241-SY-101 jet initiated fluid dynamics

International Nuclear Information System (INIS)

Trent, D.S.; Michener, T.E.

1994-01-01

The episodic Gas Release Events (GREs) that have characterized the behavior of Hanford tank 241-SY-101 for the past several years are thought to result from the entrapment of gases generated in the settled solids, i.e., sludge, layer of the tank. Gases consisting of about 36% hydrogen by volume, which are generated by complicated and poorly understood radiological and chemical processes, are apparently trapped in the settled solids layer until their accumulation initiates a buoyant upset of this layer, abruptly releasing large quantities of gas. Once concept for preventing the gas accumulation is to mobilize the settled materials with jet mixing. It is suggested that continual agitation of the settled solids using a mixer pump would free the gas bubbles so that they could continually escape, thus mitigating the potential for accumulation of flammable concentrations of hydrogen in the tank dome space following a GRE. A pump test is planned to evaluate the effectiveness of the jet mixing mitigation concept. The pump will circulate liquid from the upper layer of the tank, discharging it through two horizontal jets located approximately 2 1/2 ft above the tank floor. To prepare for start-up of this pump test, technical, operation, and safety questions concerning an anticipated gas release were addressed by numerical simulation using the TEMPEST computer code. Simulations of the pump initiated gas release revealed that the amount of gas that could potentially be released to the tank dome space is very sensitive to the initial conditions assumed for the amount and distribution of gas in the sludge layer. Calculations revealed that within the assumptions regarding gas distribution and content, the pump might initiate a rollover--followed by a significant gas release--if the sludge layer contains more than about 13 to 14% gas distributed with constant volume fraction

1/12-scale physical modeling experiments in support of tank 241-SY- 101 hydrogen mitigation

Energy Technology Data Exchange (ETDEWEB)

Fort, J.A.; Bamberger, J.A.; Bates, J.M.; Enderlin, C.W.; Elmore, M.R.

1993-01-01

Hanford tank 241-SY-101 is a 75-ft-dia double-shell tank that contains approximately 1.1 M gal of radioactive fuel reprocessing waste. Core samples have shown that the tank contents are separated into two main layers, a article laden supernatant liquid at the top of the tank and a more dense slurry on the bottom. Two additional layers may be present, one being a potentially thick sludge lying beneath the slurry at the bottom of the tank and the other being the crust that has formed on the surface of the supernatant liquid. The supernatant is more commonly referred to as the convective layer and the slurry as the non-convective layer. Accumulation of gas (partly hydrogen) in the non-convective layer is suspected to be the key mechanism behind the gas burp phenomena, and several mitigation schemes are being developed to encourage a more uniform gas release rate (Benegas 1992). To support the full-scale hydraulic mitigation test, scaled experiments were performed to satisfy two objectives: 1. provide an experimental database for numerical- model validation; 2. establish operating parameter values required to mobilize the settled solids and maintain the solids in suspension.

Minutes of the Tank Waste Science Panel meeting, July 20, 1990: Hanford Tank Safety Project

International Nuclear Information System (INIS)

Strachan, D.M.; Morgan, L.G.

1991-02-01

The second meeting of the Tank Waste Science Panel was held July 20, 1990. Science Panel members discussed the prioritization of various analyses to be performed on core samples from tank 101-SY, and were asked to review and comment on the draft Westinghouse Hanford Company document ''Analytical Chemistry Plan.'' They also reviewed and discussed the initial contributions to the report titled Chemical and Physical Processes in Tank 101-SY: A Preliminary Report. Science Panel members agreed that a fundamental understanding of the physical and chemical processes in the tank is essential, and strongly recommended that no remediation measures be taken until there is a better understanding of the chemical and physical phenomena that result in the episodic gas release from tank 101-SY. 1 ref

Composition and quantities of retained gas measured in Hanford waste tanks 241-U-103, S-106, BY-101, and BY-109

Energy Technology Data Exchange (ETDEWEB)

Mahoney, L.A.; Antoniak, Z.I.; Bates, J.M.

1997-12-01

This report provides the results obtained for the single-shell tanks (SSTs) sampled with the Retained Gas Sampler (RGS) during 1997: Tanks 241-U-103, 241-S-106, 241-BY-101, and 241-BY-109. The RGS is a modified version of the core sampler used at Hanford. It is designed specifically to be used in concert with the gas extraction equipment in the hot cell to capture and extrude a gas-containing waste sample in a hermetically sealed system. The four tanks represent several different types of flammable gas SSTs. Tank U-103 is on the Flammable Gas Watch List (FGWL) and is one of the highest-priority group of SSTs that show evidence of significant gas retention. Tank S-106, though not a FGWL tank, has a uniquely high barometric pressure response and continuing rapid surface level rise, indicating a large and increasing volume of retained gas. Tanks BY-101 and BY-109 are not on the FGWL but were chosen to test the effect of recent salt-well pumping on gas retention. Section 2 of this report provides an overview of the process by which retained gases in the Hanford tanks are sampled and analyzed. A detailed description of the procedure used to reduce and analyze the data is provided in Section 3. Tank-by-tank results are covered in Section 4 (with the data presented in the order in which the tanks were sampled), and an RGS system performance overview is given in Section 5. Section 6 presents conclusions from these analyses and recommendations for further research. The cited references are listed in Section 7. Appendix A describes the procedures used to extract gas and ammonia from the samples, Appendix B contains detailed laboratory data from each of the tanks, and Appendix C gives field sampling data.

Composition and quantities of retained gas measured in Hanford waste tanks 241-U-103, S-106, BY-101, and BY-109

International Nuclear Information System (INIS)

Mahoney, L.A.; Antoniak, Z.I.; Bates, J.M.

1997-12-01

This report provides the results obtained for the single-shell tanks (SSTs) sampled with the Retained Gas Sampler (RGS) during 1997: Tanks 241-U-103, 241-S-106, 241-BY-101, and 241-BY-109. The RGS is a modified version of the core sampler used at Hanford. It is designed specifically to be used in concert with the gas extraction equipment in the hot cell to capture and extrude a gas-containing waste sample in a hermetically sealed system. The four tanks represent several different types of flammable gas SSTs. Tank U-103 is on the Flammable Gas Watch List (FGWL) and is one of the highest-priority group of SSTs that show evidence of significant gas retention. Tank S-106, though not a FGWL tank, has a uniquely high barometric pressure response and continuing rapid surface level rise, indicating a large and increasing volume of retained gas. Tanks BY-101 and BY-109 are not on the FGWL but were chosen to test the effect of recent salt-well pumping on gas retention. Section 2 of this report provides an overview of the process by which retained gases in the Hanford tanks are sampled and analyzed. A detailed description of the procedure used to reduce and analyze the data is provided in Section 3. Tank-by-tank results are covered in Section 4 (with the data presented in the order in which the tanks were sampled), and an RGS system performance overview is given in Section 5. Section 6 presents conclusions from these analyses and recommendations for further research. The cited references are listed in Section 7. Appendix A describes the procedures used to extract gas and ammonia from the samples, Appendix B contains detailed laboratory data from each of the tanks, and Appendix C gives field sampling data

Thermocouple module halt failure acceptance test procedure for Tank 241-SY-101 DACS-1

International Nuclear Information System (INIS)

Ermi, A.M.

1997-01-01

The readiness of the Tank 241-SY-101 Data Acquisition and Control System (DACS-1) to provide monitoring and alarms for a halt failure of any thermocouple module will be tested during the performance of this procedure. Updated DACS-1 ''1/0 MODULE HEALTH STATUS'', ''MININ1'', and ''MININ2'' screens, which now provide indication of thermocouple module failure, will also be tested as part of this procedure

Data Observations on Double Shell Tank (DST) Flammable Gas Watch List Tank Behavior

Energy Technology Data Exchange (ETDEWEB)

HEDENGREN, D.C.

2000-09-28

This report provides the data from the retained gas sampler, void fraction instrument, ball rheometer, standard hydrogen monitoring system, and other tank data pertinent to gas retention and release behavior in the waste stored in double-shelled Flammable Gas Watch List tanks at Hanford. These include tanks 241-AN-103,241-AN-104, 241-AN-105, 241-AW-101, 241-SY-101, and 241-SY-103. The tanks and the waste they contain are described in terms of fill history and chemistry. The results of mixer pump operation and recent waste transfers and back-dilution in SY-101 are also described. In-situ measurement and monitoring systems are described and the data are summarized under the categories of thermal behavior, waste configuration and properties, gas generation and composition, gas retention and historical gas release behavior.

Decant pump assembly and controls qualification testing - test report

Energy Technology Data Exchange (ETDEWEB)

Staehr, T.W., Westinghouse Hanford

1996-05-02

This report summarizes the results of the qualification testing of the supernate decant pump and controls system to be used for in-tank sludge washing in aging waste tank AZ-101. The test was successful and all components are qualified for installation and use in the tank.

241-AZ Double-Shell Tanks (DST) Integrity Assessment Report

International Nuclear Information System (INIS)

JENSEN, C.E.

1999-01-01

This report presents the results of the integrity assessment of the 241-A2 double-shell tank farm facility located in the 200 East Area of the Hanford Site. The assessment included the design evaluation and integrity examinations of the tanks and concluded that the facility is adequately designed, is compatible with the waste, and is fit for use. Recommendations including subsequent examinations, are made to ensure the continued safe operation of the tanks

Laboratory testing in-tank sludge washing, summary letter report

International Nuclear Information System (INIS)

Norton, M.V.; Torres-Ayala, F.

1994-09-01

In-tank washing is being considered as a means of pretreating high-level radioactive waste sludges, such as neutralized current acid waste (NCAW) sludge. For this process, the contents of the tank will be allowed to settle, and the supernatant solution will be decanted and removed. A dilute sodium hydroxide/sodium nitrite wash solution will be added to the settled sludge and the tank contents will be mixed with a mixer pump system to facilitate washing of the sludge. After thorough mixing, the mixer pumps will be shut off and the solids will be allowed to re-settle. After settling, the supernatant solution will be withdrawn from the tank, and the wash cycle will be repeated several times with fresh wash solution. Core sample data of double shell tank 241-AZ-101 indicate that settling of NCAW solids may be very slow. A complicating factor is that strong thermal currents are expected to be generated from heat produced by radionuclides in the sludge layer at the bottom of the tank. Additionally, there are concerns that during the settling period (i.e., while mixing pumps and air-lift re-circulators are shut off), the radionuclides may heat the residual interstitial water in the sludge to the extent that violent steam discharges (steam bumping) could occur. Finally, there are concerns that during the washing steps sludge settling may be hindered as a result of the reduced ionic strength of the wash solution. To overcome the postulated reduced settling rates during the second and third washing steps, the use of flocculants is being considered. To address the above concerns and uncertainties associated with in-tank washing, PNL has conducted laboratory testing with simulant tank waste to investigate settling rates, steam bump potential, and the need for and use of flocculating agents

Supporting design information for portable exhauster installation at tanks S-109, SX-102/103, BY-105/106, S-101/102, and S-107

International Nuclear Information System (INIS)

Keller, C.M.

1997-01-01

This document provides supporting calculations and equipment dedication plans for portable exhausters and ductwork installed on tanks S-109, SX-102/103, BY-105/106, S-101/102, and S-107. The exhausters will ventilate the tanks during saltwell pumping to prevent the potential accumulation of flammable gases

Assessment of gas accumulation and retention -- Tank 241-SY-101

International Nuclear Information System (INIS)

Alleman, R.T.; Burke, T.M.; Reynolds, D.A.; Simpson, D.E.

1993-03-01

An approximate analysis has been carried out to assess and estimate the maximum quantity of gas that is likely to be accumulated within waste tank 241-SY-101, and the maximum quantity which is likely to be retained after gas release events (GRE). According to the phenomenological models used for this assessment, based on interpretation of current and recent operational data, the estimated gas generation rate in the tank is approximately 4 m 3 /day (147 ft 3 /day). About half of this gas is released as it is generated, which is (essentially) continuously. The remainder is accumulated within the slurry layer of settled solids at the bottom of the tank, and released episodically in GREs, known as ''burps,'' that are induced by unstable buoyant conditions which develop when sufficient gas accumulates in the slurry. Calculations based on gas volumes to cause neutral buoyancy in the slurry predict the following: the maximum gas accumulation (at 1 atm pressure) that can occur without triggering a GRE is in the range of 606 to 1,039 m 3 (21,400 to 36,700 ft 3 ); and the maximum gas retention immediately after a GRE is equal to the maximum accumulation minus the gas released in the GRE. GREs do not necessarily involve all of the slurry. In the largest GREs, which are assumed to involve all of the slurry, the minimum gas release (at 1 atm pressure) is calculated to be in the range of 193 to 328 m 3 (6,800 to 11,600 ft 3 ). The corresponding maximum gas retention would be 413 to 711 m 3 (14,600 to 25,100 ft 3 )

Polychlorinated biphenyls (PCB) analysis report for solid sample from 219S tank 101

International Nuclear Information System (INIS)

Diaz, L.A.

1998-01-01

One waste sample that was obtained with solids from tank 101 of 219S via a peristaltic pump equipped with a stainless steel tube and Norprene tubing (Phthalate free) was obtained in a glass jar with teflon lid was analyzed (with duplicate, matrix spike, and matrix spike duplicate) for PCBs as Aroclor mixtures by the Inorganic/Organic Chemistry Group. A soxhlet extraction procedure was used for extraction of the Aroclors from the sample. Analysis was performed using dual column confirmation gas chromatography/electron capture detection (GC/ECD). Results are presented

Mixing Envelope D Sludge with LAW Intermediate Products with and without Glass Formers

Energy Technology Data Exchange (ETDEWEB)

Hansen, E.K.

2001-09-21

The Department of Energy (DOE) Office of River Protection is in the process of designing a waste treatment system to process the Hanford Reservation High Level Waste (HLW). Envelope D sludge slurries will be blended with the concentrated Cs/Ts eluates, and the Sr/TRU intermediates separated from Envelope A, B, and C feeds. This study produced two washed simulated sludges (representing tanks 241-AZ-101 and 241-AZ-102 sludge), a Sr/TRU washed precipitate produced from tank 241-AN-107 simulant, and a concentrated blended eluate simulant based upon eluates from processing 241-AZ-102 supernate.

Inorganic, radioisotopic and organic analysis of 241-AP-101 tank waste

International Nuclear Information System (INIS)

SK Fiskum; PR Bredt; JA Campbell; LR Greenwood; OT Farmer; GJ Lumetta; GM Mong; RT Ratner; CZ Soderquist; RG Swoboda; MW Urie; JJ Wagner

2000-01-01

Battelle received five samples from Hanford waste tank 241-AP-101, taken at five different depths within the tank. No visible solids or organic layer were observed in the individual samples. Individual sample densities were measured, then the five samples were mixed together to provide a single composite. The composite was homogenized and representative sub-samples taken for inorganic, radioisotopic, and organic analysis. All analyses were performed on triplicate sub-samples of the composite material. The sample composite did not contain visible solids or an organic layer. A subsample held at 10 C for seven days formed no visible solids. The characterization of the 241-AP-101 composite samples included: (1) Inductively-coupled plasma spectrometry for Ag, Al, Ba, Bi, Ca, Cd, Cr, Cu, Fe, K, La, Mg, Mn, Na, Nd, Ni, P, Pb, Pd, Ru, Rh, Si, Sr, Ti, U, Zn, and Zr (Note: Although not specified in the test plan, As, B, Be, Co, Li, Mo, Sb, Se, Sn, Tl, V, W, and Y were also measured and reported for information only) (2) Radioisotopic analyses for total alpha and total beta activities, 3 H, 14 C, 60 Co, 79 Se, 90 Sr, 99 Tc as pertechnetate, 106 Ru/Rh, 125 Sb, 134 Cs, 137 Cs, 152 Eu, 154 Eu, 155 Eu, 238 Pu, 239+240 Pu, 241 Am, 242 Cm, and 243+244 Cm; (3) Inductively-coupled plasma mass spectrometry for 237 Np, 239 Pu, 240 Pu, 99 Tc, 126 Sn, 129 I, 231 Pa, 233 U, 234 U, 235 U, 236 U, 238 U, 241 AMU, 242 AMU, 243 AMU, As, B, Be, Ce, Co, Cs, Eu, I, Li, Mo, Pr, Rb, Sb, Se, Ta, Te, Th, Tl, V, and W; (4) total U by kinetic phosphorescence analysis; (5) Ion chromatography for Cl, F, NO 2 , NO 3 , PO 4 , SO 4 , acetate, formate, oxalate, and citrate; (6) Density, inorganic carbon and organic carbon by two different methods, mercury, free hydroxide, ammonia, and cyanide. The 241-AP-101 composite met all contract limits (molar ratio of analyte to sodium or ratio of becquerels of analyte to moles of sodium) defined in Specification 7 for Envelope A. Except for a few cases, the

1/12-scale physical modeling experiments in support of tank 241-SY- 101 hydrogen mitigation. Final report

Energy Technology Data Exchange (ETDEWEB)

Fort, J.A.; Bamberger, J.A.; Bates, J.M.; Enderlin, C.W.; Elmore, M.R.

1993-01-01

Hanford tank 241-SY-101 is a 75-ft-dia double-shell tank that contains approximately 1.1 M gal of radioactive fuel reprocessing waste. Core samples have shown that the tank contents are separated into two main layers, a article laden supernatant liquid at the top of the tank and a more dense slurry on the bottom. Two additional layers may be present, one being a potentially thick sludge lying beneath the slurry at the bottom of the tank and the other being the crust that has formed on the surface of the supernatant liquid. The supernatant is more commonly referred to as the convective layer and the slurry as the non-convective layer. Accumulation of gas (partly hydrogen) in the non-convective layer is suspected to be the key mechanism behind the gas burp phenomena, and several mitigation schemes are being developed to encourage a more uniform gas release rate (Benegas 1992). To support the full-scale hydraulic mitigation test, scaled experiments were performed to satisfy two objectives: 1. provide an experimental database for numerical- model validation; 2. establish operating parameter values required to mobilize the settled solids and maintain the solids in suspension.

Water washes and caustic leaches of sludge from Hanford Tank S-101 and water washes of sludge from Hanford Tank C-103

International Nuclear Information System (INIS)

Hunt, R.D.; Collins, J.L.; Chase, C.W.

1998-07-01

In 1993, the Department of Energy (DOE) selected the enhanced sludge washing (ESW) process as the baseline for pretreatment of Hanford tank sludges. The ESW process uses a series of water washes and caustic leaches to separate nonradioactive components such as aluminum, chromium, and phosphate from the high-level waste sludges. If the ESW process is successful, the volume of immobilized high-level waste will be significantly reduced. The tests on the sludge from Hanford Tank S-101 focused on the effects of process variables such as sodium hydroxide concentration (1 and 3 M), temperature (70 and 95 C), and leaching time (5, 24, 72, and 168 h) on the efficacy of the ESW process with realistic liquid-to-solid ratios. Another goal of this study was to evaluate the effectiveness of water washes on a sludge sample from hanford Tank C-103. The final objective of this study was to test potential process control monitors during the water washes and caustic leaches with actual sludge. Both 137 Cs activity and conductance were measured for each of the water washes and caustic leaches. Experimental procedures, a discussion of results, conclusions and recommendations are included in this report

Tank characterization report for double-shell tank 241-AN-102

International Nuclear Information System (INIS)

Jo, J.

1996-01-01

This characterization report summarizes the available information on the historical uses, current status, and sampling and analysis results of waste stored in double-shell underground storage tank 241- AN-102. This report supports the requirements of the Hanford Federal Facility Agreement and Consent Order, Milestone M-44-09 (Ecology et al. 1996). Tank 241-AN-102 is one of seven double-shell tanks located in the AN Tank Farm in the Hanford Site 200 East Area. The tank was hydrotested in 1981, and when the water was removed, a 6-inch heel was left. Tank 241-AN-102 began receiving waste from tank 241-SY-102 beginning in 1982. The tank was nearly emptied in the third quarter of 1983, leaving only 125 kL (33 kgal) of waste. Between the fourth quarter of 1983 and the first quarter of 1984, tank 241-AN-102 received waste from tanks 241-AY-102, 241-SY-102, 241-AW-105, and 241- AN-101. The tank was nearly emptied in the second quarter of 1984, leaving a heel of 129 kL (34 kgal). During the second and third quarters of 1984, the tank was filled with concentrated complexant waste from tank 241-AW-101. Since that time, only minor amounts of Plutonium-Uranium Extraction (PUREX) Plant miscellaneous waste and water have been received; there have been no waste transfer to or from the tank since 1992. Therefore, the waste currently in the tank is considered to be concentrated complexant waste. Tank 241-AN-102 is sound and is not included on any of the Watch Lists

Safety assessment for proposed pump mixing operations to mitigate episodic gas releases in tank 241-101-SY: Hanford Site, Richland, Washington

Energy Technology Data Exchange (ETDEWEB)

Lentsch, J.W., Westinghouse Hanford

1996-05-16

This safety assessment addresses each of the elements required for the proposed action to remove a slurry distributor and to install, operate, and remove a mixing pump in Tank 241-SY-101, which is located within the Hanford Site, Richland, Washington. The proposed action is required as part of an ongoing evaluation of various mitigation concepts developed to eliminate episodic gas releases that result in hydrogen concentrations in the tank dome space that exceed the lower flammability limit.

Steady-State Flammable Gas Release Rate Calculation And Lower Flammability Level Evaluation For Hanford Tank Waste

International Nuclear Information System (INIS)

Meacham, J.E.

2008-01-01

This report assesses the steady state flammability level under off normal ventilation conditions in the tank headspace for 28 double-shell tanks (DST) and 149 single shell-tanks (SST) at the Hanford Site. Flammability was calculated using estimated gas release rates, Le Chatelier's rule, and lower flammability limits of fuels in an air mixture. This revision updates the hydrogen generation rate input data for al1 177 tanks using waste composition information from the Best Basis Inventory Detail Report (data effective as of August 4,2008). Assuming only barometric breathing, the shortest time to reach 25% of the lower flammability limit is 13 days for DSTs (i.e., tank 241-AZ-102) and 36 days for SSTs (i.e., tank 241-B-203). Assuming zero ventilation, the shortest time to reach 25% of the lower flammability limit is 12 days for DSTs (i.e., tank 241-AZ-102) and 34 days for SSTs (i.e., tank 241-B-203).

FINAL REPORT INTEGRATED DM1200 MELTER TESTING OF REDOX EFFECTS USING HLW AZ-101 AND C-106/AY-102 SIMULANTS VSL-04R4800-1 REV 0 5/6/

Energy Technology Data Exchange (ETDEWEB)

KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; LUTZE W; BIZOT PM; CALLOW RA; BRANDYS M; KOT WK; PEGG IL

2011-12-29

This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 and C-106/AY-102 HLW simulants. The tests reported herein are a subset of three tests from a larger series of tests described in the Test Plan for the work; results from the remaining tests will be reported separately. Three nine day tests, one with AZ-101 and two with C-106/AY-102 feeds were conducted with variable amounts of added sugar to address the effects of redox. The test with AZ-101 included ruthenium spikes to also address the effects of redox on ruthenium volatility. One of tests addressed the effects of increased flow-sheet nitrate levels using C-106/AY-102 feeds. With high nitrate/nitrite feeds (such as WTP LAW feeds), reductants are required to prevent melt foaming and deleterious effects on glass production rates. Sugar is the baseline WTP reductant for this purpose. WTP HLW feeds typically have relatively low nitrate/nitrite content in comparison to the organic carbon content and, therefore, have typically not required sugar additions. However, HLW feed variability, particularly with respect to nitrate levels, may necessitate the use of sugar in some instances. The tests reported here investigate the effects of variable sugar additions to the melter feed as well as elevated nitrate levels in the waste. Variables held constant to the extent possible included melt temperature, bubbling rate, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW feeds with variable amounts of added sugar and increased nitrate levels; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and

Waste behavior analysis for tank 241-SY-103

International Nuclear Information System (INIS)

Wilkins, N.E.

1994-01-01

Tank 241-SY-103 is on the Flammable Gas Watch List. The waste in this tank behaves similarly to that in tank 241-Sy-101. Both show slurry growth and periodic surface level drops. However, the surface level drops are much smaller than those in tank 101-SY. A standard hydrogen monitoring system (SHMS) was recently installed in tank 103-SY, and waste auger samples were recently taken. This document covers the characterization results to date for the auger samples, and the behavior of the tank waste during both steady state periods and gas release events

Minutes of the Tank Waste Science Panel meeting September 13--14, 1990

International Nuclear Information System (INIS)

Strachan, D.M.; Morgan, L.G.

1991-02-01

The third meeting of the Tank Waste Science Panel was held September 13--14, 1990. Science Panel members were briefed on the August 5, 1990, gas release from tank 241-101-SY (commonly denoted 101-SY), synthetic waste experiments to investigate gas generation and crust behavior in the tank, computer simulations of the thermal behavior of the waste in the tank, and calculations of gas generation based on radiolytic chemistry in alkaline solutions. Data from tanks 103-SY and 103-AN were presented, but it was decided not to divert attention from tank 101-SY at this time by taking additional samples from 103-SY or 103-AN. Science Panel members recommended that multiple groups begin a concerted experimental effort to understand the chemical and physical mechanisms involved in the tank. The understanding, along with the tank model, can then be used to determine the effectiveness of a particular mitigation method before using the method in tank 101-SY. 1 tab

Mitigation of Tank 241-SY-101 by pump mixing: Results of testing phases A and B

Energy Technology Data Exchange (ETDEWEB)

Allemann, R.T.; Antoniak, Z.I.; Chvala, W.D.; Friley, J.R.; Gregory, W.B.; Hudson, J.D.; Michener, T.E.; Panisko, F.E.; Stewart, C.W.; Wise, B.M. [Pacific Northwest Lab., Richland, WA (United States); Efferding, L.E.; Fadeff, J.G.; Irwin, J.J.; Kirch, N.W. [Westinghouse Hanford Co., Richland, WA (United States)

1994-03-01

A spare mixing pump from the Hanford Grout Program was installed in Hanford double-shell waste Tank 241-SY-101 on July 3, 1993, after being modified to take advantage of waste stratification. It was anticipated that pump mixing would prevent large episodic flammable gas releases that had been occurring about every 100-150 days. A cautious initial test plan, called Phase A, was run to find how the pump and tank would behave in response to very brief and gentle pump operation. No large gas releases were triggered, and the pump performed well except for two incidents of nozzle plugging. On October 21, 1993, the next test series, Phase B, began, and the pump was applied more aggressively to mix the tank contents and mitigate uncontrolled gas releases. Orienting the pump in new directions released large volumes of gas and reduced the waste level to a near-record low. Results of the entire period from pump installation to the end of Phase B on December 17, 1993, are presented in detail in this document. Though long-term effects require further evaluation, we conclude from these data that the jet mixer pump is an effective means of controlling flammable gas release and that it has met the success criteria for mitigation in this tank.

Final Report Integrated DM1200 Melter Testing Of Redox Effects Using HLW AZ-101 And C-106/AY-102 Simulants VSL-04R4800-1, Rev. 0, 5/6/04

International Nuclear Information System (INIS)

Kruger, A.A.; Matlack, K.S.; Gong, W.; Bardakci, T.; D'Angelo, N.A.; Lutze, W.; Bizot, P.M.; Callow, R.A.; Brandys, M.; Kot, W.K.; Pegg, I.L.

2011-01-01

This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 and C-106/AY-102 HLW simulants. The tests reported herein are a subset of three tests from a larger series of tests described in the Test Plan for the work; results from the remaining tests will be reported separately. Three nine day tests, one with AZ-101 and two with C-106/AY-102 feeds were conducted with variable amounts of added sugar to address the effects of redox. The test with AZ-101 included ruthenium spikes to also address the effects of redox on ruthenium volatility. One of tests addressed the effects of increased flow-sheet nitrate levels using C-106/AY-102 feeds. With high nitrate/nitrite feeds (such as WTP LAW feeds), reductants are required to prevent melt foaming and deleterious effects on glass production rates. Sugar is the baseline WTP reductant for this purpose. WTP HLW feeds typically have relatively low nitrate/nitrite content in comparison to the organic carbon content and, therefore, have typically not required sugar additions. However, HLW feed variability, particularly with respect to nitrate levels, may necessitate the use of sugar in some instances. The tests reported here investigate the effects of variable sugar additions to the melter feed as well as elevated nitrate levels in the waste. Variables held constant to the extent possible included melt temperature, bubbling rate, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW feeds with variable amounts of added sugar and increased nitrate levels; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and

A safety assessment for proposed pump mixing operations to mitigate episodic gas releases in tank 241-SY-101: Hanford Site,Richland, Washington

Energy Technology Data Exchange (ETDEWEB)

Lentsch, J.W.

1996-07-01

This safety assessment addresses each of the elements required for the proposed action to remove a slurry distributor and to install, operate, and remove a mixing pump in Tank 241-SY-101,which is located within the Hanford Site, Richland, Washington.The proposed action is required as part of an ongoing evaluation of various mitigation concepts developed to eliminate episodic gas releases that result in hydrogen concentrations in the tank dome space that exceed the lower flammability limit.

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

Revised final report for tank 241-AN-101, grab samples 1AN-95-1 through 1AN-95-7. Revision 1

International Nuclear Information System (INIS)

Esch, R.A.

1996-01-01

Six supernate grab samples and one field blank were taken from tank 241-AN-101. This report documents analyses performed in support of the Safety Screening program: differential scanning calorimetry (DSC), thermogravimetric analysis (TGA), density by specific gravity (Sp.G.), and total alpha activity (AT)

W-030, AY/AZ tank farm cooling and miscellaneous instrumentation

International Nuclear Information System (INIS)

Cole, D.B.

1996-01-01

This is the acceptance test report for construction functional testing of Project W-030 cooling systems and related instrumentation. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The Tank Farm Cooling System consists of four forced draft cooling towers, a chilled water system, and associated controls

Decision Document for the Low Activity Waste Retrieval Strategy for Tanks 241-AN-103 and 241-AN-104 and 241-AN-105 and 241-AW-101

International Nuclear Information System (INIS)

RASMUSSEN, O.R.

2000-01-01

This report documents the preferred approach (retrieval strategy) to prepare and transfer waste from low-activity waste source tanks containing soluble solids (Tanks 241-AN-103, 241-AN-104, 241-AN-105 and 241-AW-101) to the vitrification plant. Several opportunities to further refine the selected retrieval strategy were identified; these were recommended for follow-on studies

241-SY Tank Farm Construction Extent of Condition Review for Tank Integrity

Energy Technology Data Exchange (ETDEWEB)

Barnes, Travis J.; Boomer, Kayle D.; Gunter, Jason R.; Venetz, Theodore J.

2013-07-25

This report provides the results of an extent of condition construction history review for tanks 241-SY-101, 241-SY-102, and 241-SY-103. The construction history of the 241-SY tank farm has been reviewed to identify issues similar to those experienced during tank 241-AY-102 construction. Those issues and others impacting integrity are discussed based on information found in available construction records, using tank 241-AY-102 as the comparison benchmark. In the 241-SY tank farm, the third DST farm constructed, refractory quality and stress relief were improved, while similar tank and liner fabrication issues remained.

Waste Feed Delivery System Phase 1 Preliminary Reliability and Availability and Maintainability Analysis [SEC 1 and 2

International Nuclear Information System (INIS)

CARLSON, A.B.

1999-01-01

The document presents updated results of the preliminary reliability, availability, maintainability analysis performed for delivery of waste feed from tanks 241-AZ-101 and 241-AN-105 to British Nuclear Fuels Limited, inc. under the Tank Waste Remediation System Privatization Contract. The operational schedule delay risk is estimated and contributing factors are discussed

Waste Feed Delivery System Phase 1 Preliminary Reliability and Availability and Maintainability Analysis [SEC 1 and 2

Energy Technology Data Exchange (ETDEWEB)

CARLSON, A.B.

1999-11-11

The document presents updated results of the preliminary reliability, availability, maintainability analysis performed for delivery of waste feed from tanks 241-AZ-101 and 241-AN-105 to British Nuclear Fuels Limited, inc. under the Tank Waste Remediation System Privatization Contract. The operational schedule delay risk is estimated and contributing factors are discussed.

Hanford Tank Safety Project: Minutes of the Tank Waste Science Panel meeting, February 7--8, 1991

Energy Technology Data Exchange (ETDEWEB)

Strachan, D.M. [comp.

1991-06-01

The Tank Waste Science Panel met February 7--8, 1991, to review the latest data from the analyses of the October 24, 1990, gas release from Tank 241-SY-101 (101-SY) at Hanford; discuss the results of work being performed in support of the Hanford Tank Safety Project; and be briefed on the ferrocyanide issues included in the expanded scope of the Science Panel. The shapes of the gas release curves from the past three events are similar and correlate well with changes in waste level, but the correlation between the released volume of gas and the waste height is not as good. An analysis of the kinetics of gas generation from waste height measurements in Tank 101-SY suggests that the reaction giving rise to the gases in the tank is independent of the gas pressure and independent of the physical processes that give rise to the episodic release of the gases. Tank waste height data were also used to suggest that a floating crust formed early in the history of the tank and that the current crust is being made thicker in the eastern sector of the tank by repeated upheaval of waste slurry onto the surface. The correlation between the N{sub 2}O and N{sub 2} generated in the October release appears to be 1:1, suggesting a single mechanistic pathway. Analysis of other gas generation ratios, however, suggests that H{sub 2} and N{sub 2}O are evolved together, whereas N{sub 2} is from the air. If similar ratios are observed in planned radiolysis experiments are Argonne National Laboratory, radiolysis would appear to be generating most of the gases in Tank 101-SY. Data from analysis of synthetic waste crust using a dynamic x-ray diffractometer suggest that, in air, organics are being oxidized and liberating CO{sub 2} and NO{sub x}. Experiments at Savannah River Laboratory indicate that irradiation of solutions containing NO{sub 3} and organics can produce N{sub 2}O.

Double shell tanks emergency pumping plan

International Nuclear Information System (INIS)

Tangen, M.J.

1994-01-01

At the request of the Department of Energy (DOE), a formal plan for the emergency transfer of waste from a leaking double shell tank to a designated receiver tank has been developed. This plan is in response to the priority 2 safety issue ''Response to a leaking double-shell tank'' in the DOE Report to Congress, 1991. The plan includes the tanks in four of the east tank farms and one of the west farms. The background information and supporting calculations used for the creation of the emergency plan are discussed in this document. The scope of this document is all of the double shell tanks in the AN, AP, AW, AY, and SY farms. The transfer lines, flush pits, and valve pits involved in the transfer of waste between these farms are also included in the scope. Due to the storage of high heat waste, AZ farm is excluded at this time

Analysis of several hazardous conditions for large transfer and back-dilution sequences in Tank 241-SY-101

International Nuclear Information System (INIS)

CW Stewart; LA Mahoney; WB Barton

2000-01-01

The first transfer of 89 kgal of waste and back-dilution of 61 kgal of water in Hanford Tank 241-SY-101 was accomplished December 18--20, 1999. Limits were placed on the transfer and back-dilution volumes because of concerns about potential gas release, crust sinking, and degradation of mixer pump performance. Additional transfers and back-dilutions are being planned that will bring the total to 500 kgal, which should dissolve a large fraction of the solids in the tank and dilute it well beyond the point where significant gas retention can occur. This report provides the technical bases for removing the limits on transfer and back-dilution volume by evaluating the potential consequences of several postulated hazardous conditions in view of the results of the first campaign and results of additional analyses of waste behavior

Analysis of several hazardous conditions for large transfer and back-dilution sequences in Tank 241-SY-101

Energy Technology Data Exchange (ETDEWEB)

CW Stewart; LA Mahoney; WB Barton

2000-01-28

The first transfer of 89 kgal of waste and back-dilution of 61 kgal of water in Hanford Tank 241-SY-101 was accomplished December 18--20, 1999. Limits were placed on the transfer and back-dilution volumes because of concerns about potential gas release, crust sinking, and degradation of mixer pump performance. Additional transfers and back-dilutions are being planned that will bring the total to 500 kgal, which should dissolve a large fraction of the solids in the tank and dilute it well beyond the point where significant gas retention can occur. This report provides the technical bases for removing the limits on transfer and back-dilution volume by evaluating the potential consequences of several postulated hazardous conditions in view of the results of the first campaign and results of additional analyses of waste behavior.

Mitigation of tank 241-SY-101 by pump mixing: Results of full-scale testing

International Nuclear Information System (INIS)

Stewart, C.W.; Hudson, J.D.; Friley, J.R.; Panisko, F.E.; Antoniak, Z.I.; Irwin, J.J.; Fadeff, J.G.; Efferding, L.F.; Michener, T.E.; Kirch, N.W.

1994-06-01

The Full-Scale Mixer Pump Test Program was performed in Hanford Tank 241-SY-101 from February 4 to April 13, 1994, to confirm the long-term operational strategy for flammable gas mitigation and to demonstrate that mixing can control the gas release and waste level. Since its installation on July 3, 1993, the current pump, operating only a few hours per week, has proved capable of mixing the waste sufficiently to release gas continuously instead of in large episodic events. The results of Full-Scale Testing demonstrated that the pump can control gas release and waste level for long-term mitigation, and the four test sequences formed the basis for the long-term operating schedule. The last test sequence, jet penetration tests, showed that the current pump jet creates flow near the tank wall and that it can excavate portions of the bottom sludge layer if run at maximum power. Pump mixing has altered the open-quote normal close-quote configuration of the waste; most of the original nonconvective sludge has been mixed with the supernatant liquid into a mobile convective slurry that has since been maintained by gentle pump operation and does not readily return to sludge

Project W-314 specific test and evaluation plan for AZ tank farm upgrades

International Nuclear Information System (INIS)

Hays, W.H.

1998-01-01

The purpose of this Specific Test and Evaluation Plan (STEP) is to provide a detailed written plan for the systematic testing of modifications made by the addition of the SN-631 transfer line from the AZ-O1A pit to the AZ-02A pit by the W-314 Project. The STEP develops the outline for test procedures that verify the system's performance to the established Project design criteria. The STEP is a lower tier document based on the W-314 Test and Evaluation P1 an (TEP). Testing includes Validations and Verifications (e.g., Commercial Grade Item Dedication activities, etc), Factory Tests and Inspections (FTIs), installation tests and inspections, Construction Tests and Inspections (CTIs), Acceptance Test Procedures (ATPs), Pre-Operational Test Procedures (POTPs), and Operational Test Procedures (OTPs). The STEP will be utilized in conjunction with the TEP for verification and validation

Tank Characterization Report for Single-Shell Tank 241-C-104

International Nuclear Information System (INIS)

ADAMS, M.R.

2000-01-01

Interprets information about the tank answering a series of six questions covering areas such as information drivers, tank history, tank comparisons, disposal implications, data quality and quantity, and unique aspects of the tank

Mass spectrometry analysis of tank wastes at the Hanford Site

International Nuclear Information System (INIS)

Campbell, J.A.; Mong, G.M.; Clauss, S.A.

1995-01-01

Twenty-five of the 177 high-level waste storage tanks at the Hanford Site in southeastern Washington are being watched closely because of the possibility that flammable gas mixtures may be produced from the mixed wastes contained in the storage tanks. One tank in particular, Tank 241-SY-101 (Tank 101-SY), has exhibited episodic releases of flammable gas mixtures since its final filling in the early 1980s. It has been postulated that the organic compounds present in the waste may be precursors to the production of hydrogen. Mass spectrometry has proven to be an invaluable tool for the identification of organic components in wastes from Tank 101-SY and C-103. A suite of physical and chemical analyses has been performed in support of activities directed toward the resolution of an Unresolved Safety Question concerning the potential for a floating organic layer in Hanford Waste Tank 241-C-103 to sustain a pool fire. The aqueous layer underlying the floating organic material was also analyzed for organic components

Minutes of the Tank Waste Science Panel meeting, November 11--13, 1991. Hanford Tank Safety Project

Energy Technology Data Exchange (ETDEWEB)

Strachan, D.M. [comp.

1992-04-01

The sixth meeting of the Tank Waste Science Panel was held November 11--13, 1991, in Pasco and Richland, Washington. Participating scientists presented the results of recent work on various aspects of issues relating to the generation and release of gases from Tank 241-SY-101 and the presence of ferrocyanide in other tanks at Hanford. Results are discussed.

Correlation between the surface chemistry and the atmospheric corrosion of AZ31, AZ80 and AZ91D magnesium alloys

International Nuclear Information System (INIS)

Feliu, S.; Pardo, A.; Merino, M.C.; Coy, A.E.; Viejo, F.; Arrabal, R.

2009-01-01

X-ray photoelectron spectroscopy (XPS) was used in order to investigate the correlation between the surface chemistry and the atmospheric corrosion of AZ31, AZ80 and AZ91D magnesium alloys exposed to 98% relative humidity at 50 deg. C. Commercially pure magnesium, used as the reference material, revealed MgO, Mg(OH) 2 and tracers of magnesium carbonate in the air-formed film. For the AZ80 and AZ91D alloys, the amount of magnesium carbonate formed on the surface reached similar values to those of MgO and Mg(OH) 2 . A linear relation between the amount of magnesium carbonate formed on the surface and the subsequent corrosion behaviour in the humid environment was found. The AZ80 alloy revealed the highest amount of magnesium carbonate in the air-formed film and the highest atmospheric corrosion resistance, even higher than the AZ91D alloy, indicating that aluminium distribution in the alloy microstructure influenced the amount of magnesium carbonate formed.

Tank Characterization Report for Double-Shell Tank (DST) 241-AN-107

International Nuclear Information System (INIS)

ADAMS, M.R.

2000-01-01

This report interprets information about the tank answering a series of six questions covering areas such as information drivers, tank history, tank comparisons, disposal implications, data quality and quantity, and unique aspects of the tank

A Comparative Electrochemical Study of AZ31 and AZ91 Magnesium Alloy

Directory of Open Access Journals (Sweden)

S. A. Salman

2010-01-01

Full Text Available A comparative study has been carried out on AZ31 and AZ91 magnesium alloys in order to understand the electrochemical behavior in both alkaline and chloride containing solutions. The open circuit potential (OCP was examined in 1 M NaOH and 3.5 mass % NaCl solutions. AZ31 magnesium alloy shows several potential drops throughout the immersion in 1 M NaOH solution, though AZ91 does not show this phenomenon. The specimens were anodized at a constant potential of 3 V for 30 minutes at 298 K in 1 M NaOH solution. The anticorrosion behavior of the anodized specimens was better than those of nonanodized specimens. The anodized AZ91 has better corrosion resistance compared to nonanodized specimen and anodized AZ31 magnesium alloy.

Minutes of the Tank Waste Science Panel Meeting March 25--27, 1992

International Nuclear Information System (INIS)

Schutz, W.W.; Strachan, D.M.

1992-08-01

Discussions from the seventh meeting of the Tank Waste Science are presented in Colorado. The subject areas included the generation of gases in Tank 241-SY-101, the possible use of sonication as a mitigation method, and analysis for organic constituents in core samples. Results presented and discussed include: Ferrocyanides appear to be rapidly dissolved in 1M NaOH; upon standing in the laboratory at ambient conditions oxalate precipitates from simulated wastes containing HEDTA. This suggests that one of the main components in the solids in Tank 241-SY-101 is oxalate; hydrogen evolved from waste samples from Tank 241-SY-101 is five times that observed in the off gas from the tank; data suggest that mitigation of Tank 241-SY-101 will not cause a high release of dissolved N 2 O; when using a slurry for radiation studies, a portion of the generated gases is very difficult to remove. To totally recover the generated gases, the solids must first be dissolved. This result may have an impact on mitigation by mixing if the gases are not released. Using 13 C-labeled organics in thermal degradation studies has allowed researchers to illucidate much of the kinetic mechanism for the degradation of HEDTA and glycolate. In addition to some of the intermediate, more complex organic species, oxalate, formate, and CO 2 were identified; and analytic methods for organics in radioactive complex solutions such as that found in Tank 241-SY-101 have been developed and others continue to be developed

Design of second generation Hanford tank corrosion monitoring system

International Nuclear Information System (INIS)

Edgemon, G.L.

1998-01-01

small amplitude signals that are spontaneously generated by electrochemical reactions occurring at corroding or other surfaces. Laboratory studies and recent reports on field applications have reported that EN analysis is well suited for monitoring and identifying the onset of localized corrosion, and for measuring uniform corrosion rates. A two year laboratory study was started at Hanford in 1995 to provide a technical basis for using EN in Hanford nuclear waste tanks. Based on this study, a prototype system was constructed and deployed in DST 241-AZ-101 in August, 1996. Based on the successful demonstration of this prototype for more than a year, a first-generation full-scale system was designed and installed into DST 241-AN-107 in September 1997. This document summarizes the design and operational requirements of the second-generation full-scale system scheduled for deployment into 241-AY-102

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

Single-Shell Tank (SST) Retrieval Project Plan for Tank 241-C-104 Retrieval

International Nuclear Information System (INIS)

DEFIGH PRICE, C.

2000-01-01

In support of the SST Interim Closure Project, Project W-523 ''Tank 241-C-104 Waste Retrieval System'' will provide systems for retrieval and transfer of radioactive waste from tank 241-C-104 (C-104) to the DST staging tank 241-AY-101 (AY-101). At the conclusion of Project W-523, a retrieval system will have been designed and tested to meet the requirements for Acceptance of Beneficial Use and been turned over to operations. Completion of construction and operations of the C-104 retrieval system will meet the recently proposed near-term Tri-Party Agreement milestone, M-45-03F (Proposed Tri-Party Agreement change request M-45-00-01A, August, 30 2000) for demonstrating limits of retrieval technologies on sludge and hard heels in SSTs, reduce near-term storage risks associated with aging SSTs, and provide feed for the tank waste treatment plant. This Project Plan documents the methodology for managing Project W-523; formalizes responsibilities; identifies key interfaces required to complete the retrieval action; establishes the technical, cost, and schedule baselines; and identifies project organizational requirements pertaining to the engineering process such as environmental, safety, quality assurance, change control, design verification, testing, and operational turnover

Cold-Sprayed AZ91D Coating and SiC/AZ91D Composite Coatings

Directory of Open Access Journals (Sweden)

Yingying Wang

2018-03-01

Full Text Available As an emerging coating building technique, cold spraying has many advantages to elaborate Mg alloy workpieces. In this study, AZ91D coatings and AZ91D-based composite coatings were deposited using cold spraying. Coatings were prepared using different gas temperatures to obtain the available main gas temperature. Compressed air was used as the accelerating gas, and although magnesium alloy is oxidation-sensitive, AZ91D coatings with good performance were obtained. The results show that dense coatings can be fabricated until the gas temperature is higher than 500 °C. The deposition efficiency increases greatly with the gas temperature, but it is lower than 10% for all coating specimens. To analyze the effects of compressed air on AZ91D powder particles and the effects of gas temperature on coatings, the phase composition, porosity, cross-sectional microstructure, and microhardness of coatings were characterized. X-ray diffraction and oxygen content analysis clarified that no phase transformation or oxidation occurred on AZ91D powder particles during cold spraying processes with compressed air. The porosity of AZ91D coatings remained between 3.6% and 3.9%. Impact melting was found on deformed AZ91D particles when the gas temperature increased to 550 °C. As-sprayed coatings exhibit much higher microhardness than as-casted bulk magnesium, demonstrating the dense structure of cold-sprayed coatings. To study the effects of ceramic particles on cold-sprayed AZ91D coatings, 15 vol % SiC powder particles were added into the feedstock powder. Lower SiC content in the coating than in the feedstock powder means that the deposition efficiency of the SiC powder particles is lower than the deposition efficiency of AZ91D particles. The addition of SiC particles reduces the porosity and increases the microhardness of cold-sprayed AZ91D coatings. The corrosion behavior of AZ91D coating and SiC reinforced AZ91D composite coating were examined. The Si

Methods for Heel Retrieval for Tanks C-101, C-102, and C-111 at the Hanford Site - 13064

International Nuclear Information System (INIS)

Sams, T.L.; Kirch, N.W.; Reynolds, J.H.

2013-01-01

The purpose of this paper is to evaluate the prospects of using bulk waste characteristics to determine the most appropriate heel retrieval technology. If the properties of hard to remove heels can be determined before bulk retrieval, then a heel retrieval technology can be selected before bulk retrieval is complete. This would save substantially on sampling costs and would allow the deployment of the heel retrieval technology immediately after bulk retrieval. The latter would also accelerate the heel removal schedule. A number of C-farm retrievals have been fully or partially completed at the time of this writing. Thus, there is already substantial information on the success of different technologies and the composition of the heels. There is also substantial information on the waste types in each tank based on historical records. Therefore, this study will correlate the performance of technologies used so far and compare them to the known waste types in the tanks. This will be used to estimate the performance of future C Farm heel retrievals. An initial decision tree is developed and employed on tanks C-101, C-102, and C 111. An assumption of this study is that no additional characterization information would be available, before or after retrieval. Note that collecting additional information would substantially increase the probability of success. Deploying some in-situ testing technologies, such as a water lance or an in-situ Raman probe, might substantially increase the probability of successfully selecting the process conditions without having to take samples from the tanks for laboratory analysis. (authors)

Methods for Heel Retrieval for Tanks C-101, C-102, and C-111 at the Hanford Site - 13064

Energy Technology Data Exchange (ETDEWEB)

Sams, T.L.; Kirch, N.W.; Reynolds, J.H. [Washington River protection Solutions, Richland, WA 99352 (United States)

2013-07-01

The purpose of this paper is to evaluate the prospects of using bulk waste characteristics to determine the most appropriate heel retrieval technology. If the properties of hard to remove heels can be determined before bulk retrieval, then a heel retrieval technology can be selected before bulk retrieval is complete. This would save substantially on sampling costs and would allow the deployment of the heel retrieval technology immediately after bulk retrieval. The latter would also accelerate the heel removal schedule. A number of C-farm retrievals have been fully or partially completed at the time of this writing. Thus, there is already substantial information on the success of different technologies and the composition of the heels. There is also substantial information on the waste types in each tank based on historical records. Therefore, this study will correlate the performance of technologies used so far and compare them to the known waste types in the tanks. This will be used to estimate the performance of future C Farm heel retrievals. An initial decision tree is developed and employed on tanks C-101, C-102, and C 111. An assumption of this study is that no additional characterization information would be available, before or after retrieval. Note that collecting additional information would substantially increase the probability of success. Deploying some in-situ testing technologies, such as a water lance or an in-situ Raman probe, might substantially increase the probability of successfully selecting the process conditions without having to take samples from the tanks for laboratory analysis. (authors)

Methods for heel retrieval for tanks C-101, C-102, and C-111 at the Hanford Site-13064

International Nuclear Information System (INIS)

Sams, Terry L.; Kirch, N. W.; Reynolds, Jacob G.

2013-01-01

The purpose of this paper is to evaluate the prospects of using bulk waste characteristics to determine the most appropriate heel retrieval technology. If the properties of hard to remove heels can be determined before bulk retrieval, then a heel retrieval technology can be selected before bulk retrieval is complete. This would save substantially on sampling costs and would allow the deployment of the heel retrieval technology immediately after bulk retrieval. The latter would also accelerate the heel removal schedule. A number of C-farm retrievals have been fully or partially completed at the time of this writing. Thus, there is already substantial information on the success of different technologies and the composition of the heels. There is also substantial information on the waste types in each tank based on historical records. Therefore, this study will correlate the performance of technologies used so far and compare them to the known waste types in the tanks. This will be used to estimate the performance of future C Farm heel retrievals. An initial decision tree is developed and employed on tanks C-101, C-102, and C 111. An assumption of this study is that no additional characterization information would be available, before or after retrieval. Note that collecting additional information would substantially increase the probability of success. Deploying some in-situ testing technologies, such as a water lance or an in-situ Raman probe, might substantially increase the probability of successfully selecting the process conditions without having to take samples from the tanks for laboratory analysis

TWRS privatization support project waste characterization database development

International Nuclear Information System (INIS)

1995-11-01

Pacific Northwest National Laboratory requested support from ICF Kaiser Hanford Company in assembling radionuclide and chemical analyte sample data and inventory estimates for fourteen Hanford underground storage tanks: 241-AN-102, -104, -105, -106, and -107, 241-AP-102, -104, and -105, 241-AW-101, -103, and -105, 241 AZ-101 and -102; and 241-C-109. Sample data were assembled for sixteen radionuclides and thirty-five chemical analytes. The characterization data were provided to Pacific Northwest National Laboratory in support of the Tank Waste Remediation Services Privatization Support Project. The purpose of this report is to present the results and document the methodology used in preparing the waste characterization information data set to support the Tank Waste Remediation Services Privatization Support Project. This report describes the methodology used in assembling the waste characterization information and how that information was validated by a panel of independent technical reviewers. Also, contained in this report are the various data sets created: the master data set, a subset, and an unreviewed data set. The master data set contains waste composition information for Tanks 241-AN-102 and -107, 241-AP-102 and -105, 241-AW-101; and 241-AZ-101 and -102. The subset contains only the validated analytical sample data from the master data set. The unreviewed data set contains all collected but unreviewed sample data for Tanks 241-AN-104, -105, and -106; 241-AP-104; 241-AW-103 and-105; and 241-C-109. The methodology used to review the waste characterization information was found to be an accurate, useful way to separate the invalid or questionable data from the more reliable data. In the future, this methodology should be considered when validating waste characterization information

Safe interim storage of Hanford tank wastes, draft environmental impact statement, Hanford Site, Richland, Washington

Energy Technology Data Exchange (ETDEWEB)

1994-07-01

This Draft EIS is prepared pursuant to the National Environmental Policy Act (NEPA) and the Washington State Environmental Policy Act (SEPA). DOE and Ecology have identified the need to resolve near-term tank safety issues associated with Watchlist tanks as identified pursuant to Public Law (P.L.) 101-510, Section 3137, ``Safety Measures for Waste Tanks at Hanford Nuclear Reservation,`` of the National Defense Authorization Act for Fiscal Year 1991, while continuing to provide safe storage for other Hanford wastes. This would be an interim action pending other actions that could be taken to convert waste to a more stable form based on decisions resulting from the Tank Waste Remediation System (TWRS) EIS. The purpose for this action is to resolve safety issues concerning the generation of unacceptable levels of hydrogen in two Watchlist tanks, 101-SY and 103-SY. Retrieving waste in dilute form from Tanks 101-SY and 103-SY, hydrogen-generating Watchlist double shell tanks (DSTs) in the 200 West Area, and storage in new tanks is the preferred alternative for resolution of the hydrogen safety issues.

Safe interim storage of Hanford tank wastes, draft environmental impact statement, Hanford Site, Richland, Washington

International Nuclear Information System (INIS)

1994-07-01

This Draft EIS is prepared pursuant to the National Environmental Policy Act (NEPA) and the Washington State Environmental Policy Act (SEPA). DOE and Ecology have identified the need to resolve near-term tank safety issues associated with Watchlist tanks as identified pursuant to Public Law (P.L.) 101-510, Section 3137, ''Safety Measures for Waste Tanks at Hanford Nuclear Reservation,'' of the National Defense Authorization Act for Fiscal Year 1991, while continuing to provide safe storage for other Hanford wastes. This would be an interim action pending other actions that could be taken to convert waste to a more stable form based on decisions resulting from the Tank Waste Remediation System (TWRS) EIS. The purpose for this action is to resolve safety issues concerning the generation of unacceptable levels of hydrogen in two Watchlist tanks, 101-SY and 103-SY. Retrieving waste in dilute form from Tanks 101-SY and 103-SY, hydrogen-generating Watchlist double shell tanks (DSTs) in the 200 West Area, and storage in new tanks is the preferred alternative for resolution of the hydrogen safety issues

Hanford Site Waste Storage Tank Information Notebook

International Nuclear Information System (INIS)

Husa, E.I.; Raymond, R.E.; Welty, R.K.; Griffith, S.M.; Hanlon, B.M.; Rios, R.R.; Vermeulen, N.J.

1993-07-01

This report provides summary data on the radioactive waste stored in underground tanks in the 200 East and West Areas at the Hanford Site. The summary data covers each of the existing 161 Series 100 underground waste storage tanks (500,000 gallons and larger). It also contains information on the design and construction of these tanks. The information in this report is derived from existing reports that document the status of the tanks and their materials. This report also contains interior, surface photographs of each of the 54 Watch List tanks, which are those tanks identified as Priority I Hanford Site Tank Farm Safety Issues in accordance with Public Law 101-510, Section 3137*

Implications of access hole size on tank waste retrieval system design and cost

International Nuclear Information System (INIS)

Babcock, S.M.; Kwon, D.S.; Burks, B.L.; Stoughton, R.S.; Evans, M.S.

1994-05-01

The DOE Environmental Restoration and Waste Management Robotics Technology Development Program has been investigating the application of robotics technology to the retrieval of waste from single-shell storage tanks for several years. The use of a large, ''long-reach'' manipulator to position and orient a variety of tools and other equipment has been recommended. The objective of this study is to determine the appropriate access hole size for the tank waste retrieval system installation. Previous reports on the impact of access hole size on manipulator performance are summarized. In addition, the practical limitation for access hole size based on structural limitations of the waste storage tanks, the state-of-the-art size limitations for the installation of new risers, the radiation safety implications of various access hole sizes, and overall system cost implications are considered. Basic conclusions include: (1) overall cost of remediation will; be dominated by the costs of the balance of plant and time required to perform the task rather than the cost of manipulator hardware or the cost of installing a riser, (2) the most desirable solution from a manipulator controls point of view is to make the manipulator as stiff as possible and have as high as possible a natural frequency, which implies a large access hole diameter, (3) beyond some diameter; simple, uniform cross-section elements become less advantageous from a weight standpoint and alternative structures should be considered, and (4) additional shielding and contamination control measures would be required for larger holes. Parametric studies summarized in this report considered 3,790,000 1 (1,000,000 gal) tanks, while initial applications are likely to be for 2,840,000 1 (750,000 gal) tanks. Therefore, the calculations should be somewhat conservative, recognizing the limitations of the specific conditions considered

Design review report for the SY-101 RAPID mitigation system

International Nuclear Information System (INIS)

SCHLOSSER, R.L.

1999-01-01

This report documents design reviews conducted of the SY-101 Respond And Pump In Days (RAPID) Mitigation System. As part of the SY-101 Surface-Level-Rise Remediation Project, the SY-101 WID Mitigation System will reduce the potential unacceptable consequences of crust growth in Tank 241-SY-101 (SY-101). Projections of the crust growth rate indicate that the waste level in the tank may reach the juncture of the primary and secondary confinement structures of the tank late in 1999. Because of this time constraint, many design activities are being conducted in parallel and design reviews were conducted for system adequacy as well as design implementation throughout the process. Design implementation, as used in this design review report, is the final component selection (e.g., which circuit breaker, valve, or thermocouple) that meets the approved design requirements, system design, and design and procurement specifications. Design implementation includes the necessary analysis, testing, verification, and qualification to demonstrate compliance with the system design and design requirements. Design implementation is outside the scope of this design review. The design activities performed prior to detailed design implementation (i.e., system mission requirements, functional design requirements, technical criteria, system conceptual design, and where design and build contracts were placed, the procurement specification) have been reviewed and are within the scope of this design review report. Detailed design implementation will be controlled, reviewed, and where appropriate, approved in accordance with Tank Waste Remediation System (TWRS) engineering procedures. Review of detailed design implementation will continue until all components necessary to perform the transfer function are installed and tested

Design review report for the SY-101 RAPID mitigation system

Energy Technology Data Exchange (ETDEWEB)

SCHLOSSER, R.L.

1999-05-24

This report documents design reviews conducted of the SY-101 Respond And Pump In Days (RAPID) Mitigation System. As part of the SY-101 Surface-Level-Rise Remediation Project, the SY-101 WID Mitigation System will reduce the potential unacceptable consequences of crust growth in Tank 241-SY-101 (SY-101). Projections of the crust growth rate indicate that the waste level in the tank may reach the juncture of the primary and secondary confinement structures of the tank late in 1999. Because of this time constraint, many design activities are being conducted in parallel and design reviews were conducted for system adequacy as well as design implementation throughout the process. Design implementation, as used in this design review report, is the final component selection (e.g., which circuit breaker, valve, or thermocouple) that meets the approved design requirements, system design, and design and procurement specifications. Design implementation includes the necessary analysis, testing, verification, and qualification to demonstrate compliance with the system design and design requirements. Design implementation is outside the scope of this design review. The design activities performed prior to detailed design implementation (i.e., system mission requirements, functional design requirements, technical criteria, system conceptual design, and where design and build contracts were placed, the procurement specification) have been reviewed and are within the scope of this design review report. Detailed design implementation will be controlled, reviewed, and where appropriate, approved in accordance with Tank Waste Remediation System (TWRS) engineering procedures. Review of detailed design implementation will continue until all components necessary to perform the transfer function are installed and tested.

Hanford Tank Farm interim storage phase probabilistic risk assessment outline

Energy Technology Data Exchange (ETDEWEB)

1994-05-19

This report is the second in a series examining the risks for the high level waste (HLW) storage facilities at the Hanford Site. The first phase of the HTF PSA effort addressed risks from Tank 101-SY, only. Tank 101-SY was selected as the initial focus of the PSA because of its propensity to periodically release (burp) a mixture of flammable and toxic gases. This report expands the evaluation of Tank 101-SY to all 177 storage tanks. The 177 tanks are arranged into 18 farms and contain the HLW accumulated over 50 years of weapons material production work. A centerpiece of the remediation activity is the effort toward developing a permanent method for disposing of the HLW tank`s highly radioactive contents. One approach to risk based prioritization is to perform a PSA for the whole HLW tank farm complex to identify the highest risk tanks so that remediation planners and managers will have a more rational basis for allocating limited funds to the more critical areas. Section 3 presents the qualitative identification of generic initiators that could threaten to produce releases from one or more tanks. In section 4 a detailed accident sequence model is developed for each initiating event group. Section 5 defines the release categories to which the scenarios are assigned in the accident sequence model and presents analyses of the airborne and liquid source terms resulting from different release scenarios. The conditional consequences measured by worker or public exposure to radionuclides or hazardous chemicals and economic costs of cleanup and repair are analyzed in section 6. The results from all the previous sections are integrated to produce unconditional risk curves in frequency of exceedance format.

The effect of dilution on the gas retention behavior of Tank 241-SY- 103 waste

International Nuclear Information System (INIS)

Bredt, P.R.; Tingey, S.M.

1996-01-01

Twenty-five of the 177 underground waste storage tanks on the Hanford Site have been placed on the Flammable Gas watch list. These 25 tanks, containing high-level waste generated during plutonium and uranium processing, have been identified as potentially capable of accumulating flammable gases above the lower flammability limit (Babad et al. 1991). In the case of Tanks 241-SY-101 and 241-SY-103, it has been proposed that diluting the tank waste may mitigate this hazard (Hudson et al. 1995; Stewart et al. 1994). The effect of dilution on the ability of waste from Tank 241-SY-103 to accumulate gas was studied at Pacific Northwest National Laboratory. A similar study has been completed for waste from Tank 241-SY-101 (Bredt et al. 1995). Because of the additional waste-storage volume available in Tank 241-SY-103 and because the waste is assumed to be similar to that currently in Tank 241-SY-101, Tank 241-SY-103 became the target for a demonstration of passive mitigation through in-tank dilution. In 1994, plans for the in-tank dilution demonstration were deferred pending a decision on whether to pursue dilution as a mitigation strategy. However, because Tank 241-SY-103 is an early retrieval target, determination of how waste properties vary with dilution will still be required

Gas retention and release behavior in Hanford double-shell waste tanks

Energy Technology Data Exchange (ETDEWEB)

Meyer, P.A.; Brewster, M.E.; Bryan, S.A. [and others

1997-05-01

This report describes the current understanding of flammable gas retention and release in Hanford double-shell waste tanks AN-103, AN-104, AN-105, AW-101, SY-101, and SY-103. This knowledge is based on analyses, experimental results, and observations of tank behavior. The applicable data available from the void fraction instrument, retained gas sampler, ball rheometer, tank characterization, and field monitoring are summarized. Retained gas volumes and void fractions are updated with these new data. Using the retained gas compositions from the retained gas sampler, peak dome pressures during a gas burn are calculated as a function of the fraction of retained gas hypothetically released instantaneously into the tank head space. Models and criteria are given for gas generation, initiation of buoyant displacement, and resulting gas release; and predictions are compared with observed tank behavior.

Gas retention and release behavior in Hanford double-shell waste tanks

International Nuclear Information System (INIS)

Meyer, P.A.; Brewster, M.E.; Bryan, S.A.

1997-05-01

This report describes the current understanding of flammable gas retention and release in Hanford double-shell waste tanks AN-103, AN-104, AN-105, AW-101, SY-101, and SY-103. This knowledge is based on analyses, experimental results, and observations of tank behavior. The applicable data available from the void fraction instrument, retained gas sampler, ball rheometer, tank characterization, and field monitoring are summarized. Retained gas volumes and void fractions are updated with these new data. Using the retained gas compositions from the retained gas sampler, peak dome pressures during a gas burn are calculated as a function of the fraction of retained gas hypothetically released instantaneously into the tank head space. Models and criteria are given for gas generation, initiation of buoyant displacement, and resulting gas release; and predictions are compared with observed tank behavior

The Effects of Rare Earth Pr and Heat Treatment on the Wear Properties of AZ91 Alloy

Directory of Open Access Journals (Sweden)

Ning Li

2018-06-01

Full Text Available This paper investigated the influences of Pr addition and heat treatment (T6 on the dry sliding wear behavior of AZ91 alloy. The wear rates and friction coefficients were measured by using a pin-on-disc tribometer under loads of 30, 60 and 90 N at dry sliding speeds of 100 rpm, over a sliding time of 15 min. The worn surfaces were examined using a scanning electron microscope and was analyzed with an energy dispersive spectrometer. The experimental results revealed that AZ91-1.0%Pr magnesium alloy exhibited lower wear rate and friction coefficient than the other investigated alloys. As the applied load increased, the wear rate and friction coefficient increased. Compared with the as-cast AZ91-1.0%Pr magnesium alloy, the hardness and wear resistance of the alloy after solution treatment were reduced, and through the subsequent aging, the hardness and wear resistance of the alloy were improved and the hardness was 101.1 HB (compared to as-cast AZ91 magnesium alloy, it increased by 45%. The AZ91-1.0%Pr with T6 magnesium alloy exhibited best wear resistance. Abrasion was dominant at load of 30 N, delamination was dominant at load of 60 N and plastic deformation was dominant at load of 90 N. Oxidation was observed at all loads.

Solid-phase characterization in flammable-gas-tank sludges by electron microscopy

International Nuclear Information System (INIS)

Liu, J.; Pederson, L.R.; Qang, L.Q.

1995-09-01

The crystallinity, morphology, chemical composition, and crystalline phases of several Tank 241-SY-101 (hereinafter referred to as SY-101) and Tank 241-SY-103 (hereinafter referred to as SY-103) solid samples were studied by transmission electron microscopy (TEM), electron energy dispersive spectroscopy (EDS), and electron diffraction. The main focus is on the identification of aluminum hydroxide thought to be present in these tank samples. Aluminum hydroxide was found in SY-103, but not in SY-101. This difference can be explained by the different OH/Al ratios found in the two tank samples: a high OH/Al ratio in SY-101 favors the formation of sodium aluminate, but a low OH/Al ratio in SY-103 favors aluminum hydroxide. These results were confirmed by a magnetic resonance study on SY-101 and SY-103 simulant. The transition from aluminum hydroxide to sodium aluminate occurs at an OH/Al molar ratio of 3.6. It is believed that the study of Al(OH) 3 was not affected by sample preparation because all Al(OH) 3 is in the solid form according to the NMR experiments. There is no Al(OH) 3 in the liquid. It is, therefore, most likely that the observation of Al(OH) 3 is representative of the real sludge sample, and is not affected by drying. Similar conclusions also apply to other insoluble phases such as iron and chromium

Tank 241-AX-101, grab samples, 1AX-97-1 through 1AX-97-3 analytical results for the final report

International Nuclear Information System (INIS)

Esch, R.A.

1997-01-01

This document is the final report for tank 241-AX-101 grab samples. Four grab samples were collected from riser 5B on July 29, 1997. Analyses were performed on samples 1AX-97-1, 1AX-97-2 and 1AX-97-3 in accordance with the Compatibility Grab Sampling and Analysis Plan (TSAP) and the Data Quality Objectives for Tank Farms Waste Compatibility Program (DQO) (Rev. 1: Fowler, 1995; Rev. 2: Mulkey and Miller, 1997). The analytical results are presented in Table 1. No notification limits were exceeded. All four samples contained settled solids that appeared to be large salt crystals that precipitated upon cooling to ambient temperature. Less than 25 % settled solids were present in the first three samples, therefore only the supernate was sampled and analyzed. Sample 1AX-97-4 contained approximately 25.3 % settled solids. Compatibility analyses were not performed on this sample. Attachment 1 is provided as a cross-reference for relating the tank farm customer identification numbers with the 222-S Laboratory sample numbers and the portion of sample analyzed. Table 2 provides the appearance information. All four samples contained settled solids that appeared to be large salt crystal that precipitated upon cooling to ambient temperature. The settled solids in samples 1AX-97-1, 1AX-97-2 and 1AX-97-3 were less than 25% by volume. Therefore, for these three samples, two 15-mL subsamples were pipetted to the surface of the liquid and submitted to the laboratory for analysis. In addition, a portion of the liquid was taken from each of the these three samples to perform an acidified ammonia analysis. No analysis was performed on the settled solid portion of the samples. Sample 1AX-97-4 was reserved for the Process Chemistry group to perform boil down and dissolution testing in accordance with Letter of Instruction for Non-Routine Analysis of Single-Shell Tank 241-AX-101 Grab Samples (Field, 1997) (Correspondence 1). However, prior to the analysis, the sample was inadvertently

Tank 241-AX-101 grab samples 1AX-97-1 through 1AX-97-3 analytical results for the final report

Energy Technology Data Exchange (ETDEWEB)

Esch, R.A.

1997-11-13

This document is the final report for tank 241-AX-101 grab samples. Four grab samples were collected from riser 5B on July 29, 1997. Analyses were performed on samples 1AX-97-1, 1AX-97-2 and 1AX-97-3 in accordance with the Compatibility Grab Sampling and Analysis Plan (TSAP) and the Data Quality Objectives for Tank Farms Waste Compatibility Program (DQO) (Rev. 1: Fowler, 1995; Rev. 2: Mulkey and Miller, 1997). The analytical results are presented in Table 1. No notification limits were exceeded. All four samples contained settled solids that appeared to be large salt crystals that precipitated upon cooling to ambient temperature. Less than 25 % settled solids were present in the first three samples, therefore only the supernate was sampled and analyzed. Sample 1AX-97-4 contained approximately 25.3 % settled solids. Compatibility analyses were not performed on this sample. Attachment 1 is provided as a cross-reference for relating the tank farm customer identification numbers with the 222-S Laboratory sample numbers and the portion of sample analyzed. Table 2 provides the appearance information. All four samples contained settled solids that appeared to be large salt crystal that precipitated upon cooling to ambient temperature. The settled solids in samples 1AX-97-1, 1AX-97-2 and 1AX-97-3 were less than 25% by volume. Therefore, for these three samples, two 15-mL subsamples were pipetted to the surface of the liquid and submitted to the laboratory for analysis. In addition, a portion of the liquid was taken from each of the these three samples to perform an acidified ammonia analysis. No analysis was performed on the settled solid portion of the samples. Sample 1AX-97-4 was reserved for the Process Chemistry group to perform boil down and dissolution testing in accordance with Letter of Instruction for Non-Routine Analysis of Single-Shell Tank 241-AX-101 Grab Samples (Field, 1997) (Correspondence 1). However, prior to the analysis, the sample was inadvertently

Qualification test for the Flexible Receiver. Revision 1

International Nuclear Information System (INIS)

Keller, C.M.

1994-01-01

This document provides the test plan and procedures to certify and design verify the 42 in. and 4 in. -- 6 in. Flexible Receiver as a safety class 3 system. The Flexible Receiver will be used by projects W-151 and W-320 for removing equipment from tanks C-106 and Az-101

Hanford Tank Farm interim storage phase probabilistic risk assessment outline

International Nuclear Information System (INIS)

1994-01-01

This report is the second in a series examining the risks for the high level waste (HLW) storage facilities at the Hanford Site. The first phase of the HTF PSA effort addressed risks from Tank 101-SY, only. Tank 101-SY was selected as the initial focus of the PSA because of its propensity to periodically release (burp) a mixture of flammable and toxic gases. This report expands the evaluation of Tank 101-SY to all 177 storage tanks. The 177 tanks are arranged into 18 farms and contain the HLW accumulated over 50 years of weapons material production work. A centerpiece of the remediation activity is the effort toward developing a permanent method for disposing of the HLW tank's highly radioactive contents. One approach to risk based prioritization is to perform a PSA for the whole HLW tank farm complex to identify the highest risk tanks so that remediation planners and managers will have a more rational basis for allocating limited funds to the more critical areas. Section 3 presents the qualitative identification of generic initiators that could threaten to produce releases from one or more tanks. In section 4 a detailed accident sequence model is developed for each initiating event group. Section 5 defines the release categories to which the scenarios are assigned in the accident sequence model and presents analyses of the airborne and liquid source terms resulting from different release scenarios. The conditional consequences measured by worker or public exposure to radionuclides or hazardous chemicals and economic costs of cleanup and repair are analyzed in section 6. The results from all the previous sections are integrated to produce unconditional risk curves in frequency of exceedance format

Preoperational test report, primary ventilation condenser cooling system

Energy Technology Data Exchange (ETDEWEB)

Clifton, F.T.

1997-10-29

This represents the preoperational test report for the Primary Ventilation Condenser Cooling System, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system uses a closed chilled water piping loop to provide offgas effluent cooling for tanks AY101, AY102, AZ1O1, AZ102; the offgas is cooled from a nominal 100 F to 40 F. Resulting condensation removes tritiated vapor from the exhaust stack stream. The piping system includes a package outdoor air-cooled water chiller with parallel redundant circulating pumps; the condenser coil is located inside a shielded ventilation equipment cell. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System.

Preoperational test report, primary ventilation condenser cooling system

International Nuclear Information System (INIS)

Clifton, F.T.

1997-01-01

This represents the preoperational test report for the Primary Ventilation Condenser Cooling System, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system uses a closed chilled water piping loop to provide offgas effluent cooling for tanks AY101, AY102, AZ1O1, AZ102; the offgas is cooled from a nominal 100 F to 40 F. Resulting condensation removes tritiated vapor from the exhaust stack stream. The piping system includes a package outdoor air-cooled water chiller with parallel redundant circulating pumps; the condenser coil is located inside a shielded ventilation equipment cell. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System

Hanford high level waste (HLW) tank mixer pump safe operating envelope reliability assessment

International Nuclear Information System (INIS)

Fischer, S.R.; Clark, J.

1993-01-01

The US Department of Energy and its contractor, Westinghouse Corp., are responsible for the management and safe storage of waste accumulated from processing defense reactor irradiated fuels for plutonium recovery at the Hanford Site. These wastes, which consist of liquids and precipitated solids, are stored in underground storage tanks pending final disposition. Currently, 23 waste tanks have been placed on a safety watch list because of their potential for generating, storing, and periodically releasing various quantities of hydrogen and other gases. Tank 101-SY in the Hanford SY Tank Farm has been found to release hydrogen concentrations greater than the lower flammable limit (LFL) during periodic gas release events. In the unlikely event that an ignition source is present during a hydrogen release, a hydrogen burn could occur with a potential to release nuclear waste materials. To mitigate the periodic gas releases occurring from Tank 101-SY, a large mixer pump currently is being installed in the tank to promote a sustained release of hydrogen gas to the tank dome space. An extensive safety analysis (SA) effort was undertaken and documented to ensure the safe operation of the mixer pump after it is installed in Tank 101-SY.1 The SA identified a need for detailed operating, alarm, and abort limits to ensure that analyzed safety limits were not exceeded during pump operations

Tank Space Options Report

International Nuclear Information System (INIS)

BOYLES, V.C.

2001-01-01

A risk-based priority for the retrieval of Hanford Site waste from the 149 single-shell tanks (SSTs) has been adopted as a result of changes to the Hanford Federal Facility Agreement and Consent Order (HFFACO) (Ecology et al. 1997) negotiated in 2000. Retrieval of the first three tanks in the retrieval sequence fills available capacity in the double-shell tanks (DSTs) by 2007. As a result, the HFFACO change established a milestone (M-45-12-TO1) requiring the determination of options that could increase waste storage capacity for single-shell tank waste retrieval. The information will be considered in future negotiations. This document fulfills the milestone requirement. This study presents options that were reviewed for the purpose of increasing waste storage capacity. Eight options are identified that have the potential for increasing capacity from 5 to 10 million gallons, thus allowing uninterrupted single-shell tank retrieval until the planned Waste Treatment Plant begins processing substantial volumes of waste from the double-shell tanks in 2009. The cost of implementing these options is estimated to range from less than $1 per gallon to more than $14 per gallon. Construction of new double-shell tanks is estimated to cost about $63 per gallon. Providing 5 to 10 million gallons of available double-shell tank space could enable early retrieval of 5 to 9 high-risk single-shell tanks beyond those identified for retrieval by 2007. These tanks are A-101, AX-101, AX-103, BY-102, C-107, S-105, S-106, S-108, and S-109 (Garfield et al. 2000). This represents a potential to retrieve approximately 14 million total curies, including 3,200 curies of long-lived mobile radionuclides. The results of the study reflect qualitative analyses conducted to identify promising options. The estimated costs are rough-order-of magnitude and, therefore, subject to change. Implementing some of the options would represent a departure from the current baseline and may adversely impact the

Radioactive Air Emissions Notice of Construction Application for Installation and Operation of a Waste Retrieval System in Tanks 241-AN-101, 241-AN--102, 241-AN--103, 241-AN--104, 241-AN--105, and 241-AN--107

International Nuclear Information System (INIS)

HILL, J.S.

2000-01-01

This document serves as a notice of construction (NOC) pursuant to the requirements of Washington Administrative Code (WAC) 246-247-060, and as a request for approval to modify pursuant to 40 Code of Federal Regulations (CFR) 61.07, for the installation and operation of one waste retrieval system in each of the following tanks; 241-AN-101, -AN-102, -AN-103, -AN-104, -AN-105 and -AN-107. Pursuant to 40 CFR 61.09 (aXI), this application is also intended to provide anticipated initial start-up notification. It is requested that EPA approval of this application will also constitute EPA acceptance of the initial start-up notification. This NOC covers the installation and operation o f a waste retrieval system in tanks 241-AN-101, -AN-102, -AN-103, -AN-104, -AN-105 and -AN-107, and the 241-AN-A/-B Valve Pits. Generally, this includes removal of existing equipment, installation of new equipment, and construction of new ancillary equipment and buildings between now and the year2011. Tanks 241-AN-101, -AN-102, -AN-103, -AN-104, -AN-105 and -AN-107 will provide waste feed for immobilization into a low activity waste (LAW) product

Qualification test for the flexible receiver. Revision 2

Energy Technology Data Exchange (ETDEWEB)

Tedeschi, D.J.

1994-12-12

This document provides the test plan and procedures to certify and design verify the 42{double_prime} and 4{double_prime}-6{double_prime} Flexible Receiver as a safety class 3 system. The Flexible Receiver will be used by projects W-151 and W-320 for removing equipment from tanks C-106 and AZ-101.

Qualification test for the flexible receiver. Revision 2

International Nuclear Information System (INIS)

Tedeschi, D.J.

1994-01-01

This document provides the test plan and procedures to certify and design verify the 42 double-prime and 4 double-prime-6 double-prime Flexible Receiver as a safety class 3 system. The Flexible Receiver will be used by projects W-151 and W-320 for removing equipment from tanks C-106 and AZ-101

Evaluation of Hanford Single-Shell Waste Tanks Suspected of Water Intrusion

International Nuclear Information System (INIS)

Feero, Amie J.; Washenfelder, Dennis J.; Johnson, Jeremy M.; Schofield, John S.

2013-01-01

Intrusions evaluations for twelve single-shell tanks were completed in 2013. The evaluations consisted of remote visual inspections, data analysis, and calculations of estimated intrusion rates. The observation of an intrusion or the preponderance of evidence confirmed that six of the twelve tanks evaluated had intrusions. These tanks were tanks 241-A-103, BX-101, BX-103, BX-110, BY-102, and SX-106

In situ rheology and gas volume in Hanford double-shell waste tanks

International Nuclear Information System (INIS)

Stewart, C.W.; Alzheimer, J.M.; Brewster, M.E.; Chen, G.; Reid, H.C.; Shepard, C.L.; Terrones, G.; Mendoza, R.E.

1996-09-01

This report is a detailed characterization of gas retention and release in 6 Hanford DS waste tanks. The results came from the ball rheometer and void fraction instrument in (flammable gas watch list) tanks SY-101, SY-103, AW-101, AN-103, AN-104, and AN-105 are presented. Instrument operation and derivation of data reduction methods are presented. Gas retention and release information is summarized for each tank and includes tank fill history and instrumentation, waste configuration, gas release, void fraction distribution, gas volumes, rheology, and photographs of the waste column from extruded core samples. Potential peak burn pressure is computed as a function of gas release fraction to portray the 'hazard signature' of each tank. It is shown that two tanks remain well below the maximum allowable pressure, even if the entire gas content were released and ignited, and that none of the others present a hazard with their present gas release behavior

Slurry growth, gas retention, and flammable gas generation by Hanford radioactive waste tanks: Synthetic waste studies, FY 1991

International Nuclear Information System (INIS)

Bryan, S.A.; Pederson, L.R.; Ryan, J.L.; Scheele, R.D.; Tingey, J.M.

1992-08-01

Of 177 high-level waste storage tanks on the Hanford Site, 23 have been placed on a safety watch list because they are suspected of producing flammable gases in flammable or explosive concentrate. One tankin particular, Tank 241-SY-101 (Tank 101-SY), has exhibited slow increases in waste volume followed by a rapid decrease accompanied by venting of large quantities of gases. The purpose of this study is to help determine the processes by which flammable gases are produced, retained, and eventually released from Tank 101-SY. Waste composition data for single- and double-shell waste tanks on the flammable gas watch listare critically reviewed. The results of laboratory studies using synthetic double-shell wastes are summarized, including physical and chemical properties of crusts that are formed, the stoichiometry and rate ofgas generation, and mechanisms responsible for formation of a floating crust

Hanford Tank Farms Vadose Zone, Addendum to the T Tank Farm Report

Energy Technology Data Exchange (ETDEWEB)

Spatz, Robert

2000-07-01

This addendum to the T Tank Farm Report (GJO-99-101-TARA, GJO-HAN-27) published in September 1999 incorporates the results of high-rate and repeat logging activities along with shape factor analysis of the logging incorporates the results of high-rate and repeat logging activities along with shape factor analysis of the logging data. A high-rate logging system was developed and deployed in the T Tank Farm to measure cesium-137 concentration levels in high gamma flux zones where the spectral gamma logging system was unable to collect usable data because of high dead times and detector saturation. This report presents additional data and revised visualizations of subsurface contaminant distribution in the T Tank Farm at the DOE Hanford Site in the state of Washington.

ELECTROCHEMICAL STUDIES OF CARBON STEEL CORROSION IN HANFORD DOUBLE SHELL TANK (DST) WASTE

Energy Technology Data Exchange (ETDEWEB)

DUNCAN, J.B.; WINDISCH, C.F.

2006-10-13

This paper reports on the electrochemical scans for the supernatant of Hanford double-shell tank (DST) 241-SY-102 and the electrochemical scans for the bottom saltcake layer for Hanford DST 241-AZ-102. It further reports on the development of electrochemical test cells adapted to both sample volume and hot cell constraints.

A comparative corrosion behavior of Mg, AZ31 and AZ91 alloys in 3.5% NaCl solution

Directory of Open Access Journals (Sweden)

I.B. Singh

2015-06-01

Full Text Available The corrosion behavior of Mg, AZ31 and AZ91 has been evaluated in 3.5% NaCl solution using weight loss, electrochemical polarization and impedance measurements. Corrosion rate derived from the weight losses demonstrated the occurrence of steeply fast corrosion reaction on AZ91 alloy after three hours of immersion, indicating the start of galvanic corrosion. An increase of corrosion rate with immersion time was also observed for AZ31 but with lesser extent than AZ91 alloy. Whereas Mg metals showed a decrease of corrosion rate with immersion time, suggesting the formation of a protective layer on their surfaces. In contrast, the corrosion current density (Icorr derived from the Tafel plots, exhibited their corrosion resistances in order of Mg > AZ91 > AZ31. Electrochemical charge transfer resistance (Rct and double layer capacitance measured by electrochemical impedance spectroscopy (EIS, are well in accordance with the measured Icorr. EIS measurements with time and microstructural examination of the corroded and uncorroded samples are helpful in elucidation of results measured by electrochemical polarization.

Minutes of the Tank Waste Science Panel meeting, November 11--13, 1991

Energy Technology Data Exchange (ETDEWEB)

Strachan, D.M. (comp.)

1992-04-01

The sixth meeting of the Tank Waste Science Panel was held November 11--13, 1991, in Pasco and Richland, Washington. Participating scientists presented the results of recent work on various aspects of issues relating to the generation and release of gases from Tank 241-SY-101 and the presence of ferrocyanide in other tanks at Hanford. Results are discussed.

Minutes of the Tank Waste Science Panel meeting, November 11--13, 1991

International Nuclear Information System (INIS)

Strachan, D.M.

1992-04-01

The sixth meeting of the Tank Waste Science Panel was held November 11--13, 1991, in Pasco and Richland, Washington. Participating scientists presented the results of recent work on various aspects of issues relating to the generation and release of gases from Tank 241-SY-101 and the presence of ferrocyanide in other tanks at Hanford. Results are discussed

Final Report Tests On The Duramelter 1200 HLW Pilot Melter System Using AZ-101 HLW Simulants VSL-02R0100-2, Rev. 1, 2/17/03

International Nuclear Information System (INIS)

Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Bardakci, T.; Gong, W.; D'Angelo, N.A.; Schatz, T.R.; Pegg, I.L.

2011-01-01

WVDP facility, lending confidence to the tests results (1). Since the inclusion or exclusion of a bubbler has significant design implications, the Project commissioned further tests to address this issue. In an effort to identify factors that might increase the glass production rate for projected WTP melter feeds, a subsequent series of tests was performed on the DM100 system. Several tests variables led to glass production rate increases to values significantly above the 400 kg/m2/d requirement. However, while small-scale melter tests are useful for screening relative effects, they tend to overestimate absolute glass production rates, particularly for un-bubbled tests. Consequently, when scale-up effects were taken into account, it was not clear that any of the variables investigated would conclusively meet the 400 kg/m 2 /d requirement without bubbling. The present series of tests was therefore performed on the DM1200 one-third scale HLW pilot melter system to provide the required basis for a final decision on whether bubblers would be included in the HLW melter. The present tests employed the same AZ-101 waste simulant and glass composition that was used for previous testing for consistency and comparability with the results from the earlier tests.

FINAL REPORT TESTS ON THE DURAMELTER 1200 HLW PILOT MELTER SYSTEM USING AZ-101 HLW SIMULANTS VSL-02R0100-2 REV 1 2/17/03

Energy Technology Data Exchange (ETDEWEB)

KRUGER AA; MATLACK KS; KOT WK; BARDAKCI T; GONG W; D' ANGELO NA; SCHATZ TR; PEGG IL

2011-12-29

larger WVDP facility, lending confidence to the tests results [1]. Since the inclusion or exclusion of a bubbler has significant design implications, the Project commissioned further tests to address this issue. In an effort to identify factors that might increase the glass production rate for projected WTP melter feeds, a subsequent series of tests was performed on the DM100 system. Several tests variables led to glass production rate increases to values significantly above the 400 kg/m2/d requirement. However, while small-scale melter tests are useful for screening relative effects, they tend to overestimate absolute glass production rates, particularly for un-bubbled tests. Consequently, when scale-up effects were taken into account, it was not clear that any of the variables investigated would conclusively meet the 400 kg/m{sup 2}/d requirement without bubbling. The present series of tests was therefore performed on the DM1200 one-third scale HLW pilot melter system to provide the required basis for a final decision on whether bubblers would be included in the HLW melter. The present tests employed the same AZ-101 waste simulant and glass composition that was used for previous testing for consistency and comparability with the results from the earlier tests.

Acceptance test procedure, 241-SY-101/241-C-106 shot loading system

International Nuclear Information System (INIS)

Ostrom, M.J.

1994-01-01

This Acceptance Test Procedure is for the 241-SY-101/241-C-106 Shot Loading System. The procedure will test the components of the Shot Loading System and its capability of adequately loading shot into the annular space of the Container. The loaded shot will provide shielding as required for transporting and storage of a contaminated pump after removal from the tank. This test serves as verification that the SLS is acceptable for use in the pump removal operations for Tanks 241-SY-101, 241-C-106 and 241-AY-102. The pump removal operation for these three tanks will be performed by two different organizations with different equipment, but the Shot Loading System will be compatible between the two operations

Minutes of the Tank Waste Science Panel meeting July 9--1, 1991

International Nuclear Information System (INIS)

Strachan, D.M.

1992-04-01

The fifth meeting of the Tank Waste Science Panel was held July 9--11, 1991, in Atlanta, Georgia. The subject areas included the generation, retention, and release of gases from Tank 241-SY-101 and the chemistry of ferrocyanide wastes

System design description for SY-101 hydrogen mitigation test project data acquisition and control system (DACS-1)

International Nuclear Information System (INIS)

Truitt, R.W.; Pounds, T.S.; Smith, S.O.

1994-01-01

This document describes the hardware subsystems of the data acquisition and control system (DACS) used in mitigation tests conducted on waste tank SY-101 at the Hanford Nuclear Reservation. The system was designed and implemented by Los Alamos National Laboratory (LANL) and supplied to Westinghouse Hanford Company (WHC). The mitigation testing uses a pump immersed in the waste tank, directed at certain angles and operated at different speeds and time durations. The SY-101 tank has experienced recurrent periodic gas releases of hydrogen, nitrous oxide, ammonia, and (recently discovered) methane. The hydrogen gas represents a danger, as some of the releases are in amounts above the lower flammability limit (LFL). These large gas releases must be mitigated. Several instruments have been added to the tank to monitor the gas compositions, the tank level, the tank temperature, and other parameters. A mixer pump has been developed to stir the tank waste to cause the gases to be released at a slow rate. It is the function of the DACS to monitor those instruments and to control the mixer pump in a safe manner. During FY93 and FY94 the mixer pump was installed with associated testing operations support equipment and a mitigation test project plan was implemented. These activities successfully demonstrated the mixer pump's ability to mitigate the SY-101 tank hydrogen gas hazard

Suspended Solids Profiler Shop Test Report

International Nuclear Information System (INIS)

STAEHR, T.W.

2000-01-01

The Suspended Solids Profiler (SSP) Instrument is planned to be installed in the AZ-101 tank to measure suspended solids concentrations during mixer pump testing. The SSP sensor uses a reflectance measurement principle to determine the suspended solids concentrations. The purpose of this test is to provide a documented means of verifying that the functional components of the SSP operate properly

Joining of AZ31 and AZ91 Mg alloys by friction stir welding

Directory of Open Access Journals (Sweden)

B. Ratna Sunil

2015-12-01

Full Text Available Two dissimilar magnesium (Mg alloy sheets, one with low aluminium (AZ31 and another with high aluminium (AZ91 content, were successfully joined by friction stir welding (FSW. The effect of process parameters on the formation of hot cracks was investigated. A sound metallurgical joint was obtained at optimized process parameters (1400 rpm with 25 mm/min feed which contained fine grains and distributed β (Mg17Al12 phase within the nugget zone. An increasing trend in the hardness measurements has also confirmed more amount of dissolution of aluminium within the nugget zone. A sharp interface between nugget zone and thermo mechanical affected zone (TMAZ was clearly noticed at the AZ31 Mg alloy side (advancing but not on the AZ91 Mg alloy side (retreating. From the results it can be concluded that FSW can be effectively used to join dissimilar metals, particularly difficult to process metals such as Mg alloys, and hot cracking can be completely eliminated by choosing appropriate process parameters to achieve sound joint.

Program plan for evaluation and remediation of the generation and release of flammable gases in Hanford Site waste tanks

International Nuclear Information System (INIS)

Johnson, G.D.

1991-08-01

This program plan describes the activities being conducted for the resolution of the flammable gas problem that is associated with 23 high-level waste tanks at the Hanford Site. The classification of the wastes in all of these tanks is not final and some wastes may not be high-level wastes. However, until the characterization and classification is complete, all the tanks are treated as if they contain high-level waste. Of the 23 tanks, Tank 241-SY-101 (referred to as Tank 101-SY) has exhibited significant episodic releases of flammable gases (hydrogen and nitrous oxide) for the past 10 years. The major near-term focus of this program is for the understanding and stabilization of this tank. An understanding of the mechanism for gas generation and the processes for the episodic release will be obtained through sampling of the tank contents, laboratory studies, and modeling of the tank behavior. Additional information will be obtained through new and upgraded instrumentation for the tank. A number of remediation, or stabilization, concepts will be evaluated for near-term (2 to 3 years) applications to Tank 101-SY. Detailed safety assessments are required for all activities that will occur in the tank (sampling, removal of equipment, and addition of new instruments). This program plan presents a discussion of each task, provides schedules for near-term activities, and gives a summary of the expected work for fiscal years 1991, 1992, and 1993. 16 refs., 7 figs., 8 tabs

Program plan for evaluation and remediation of the generation and release of flammable gases in Hanford Site waste tanks

Energy Technology Data Exchange (ETDEWEB)

Johnson, G.D. (comp.)

1991-08-01

This program plan describes the activities being conducted for the resolution of the flammable gas problem that is associated with 23 high-level waste tanks at the Hanford Site. The classification of the wastes in all of these tanks is not final and some wastes may not be high-level wastes. However, until the characterization and classification is complete, all the tanks are treated as if they contain high-level waste. Of the 23 tanks, Tank 241-SY-101 (referred to as Tank 101-SY) has exhibited significant episodic releases of flammable gases (hydrogen and nitrous oxide) for the past 10 years. The major near-term focus of this program is for the understanding and stabilization of this tank. An understanding of the mechanism for gas generation and the processes for the episodic release will be obtained through sampling of the tank contents, laboratory studies, and modeling of the tank behavior. Additional information will be obtained through new and upgraded instrumentation for the tank. A number of remediation, or stabilization, concepts will be evaluated for near-term (2 to 3 years) applications to Tank 101-SY. Detailed safety assessments are required for all activities that will occur in the tank (sampling, removal of equipment, and addition of new instruments). This program plan presents a discussion of each task, provides schedules for near-term activities, and gives a summary of the expected work for fiscal years 1991, 1992, and 1993. 16 refs., 7 figs., 8 tabs.

Study with a steam tracer in a zone near well Az-64, in the Los Azufres geothermal field, Mich.; Estudio con un trazador de vapor en la zona aledana al pozo Az-64, en el campo geotermico de Los Azufres, Mich.

Energy Technology Data Exchange (ETDEWEB)

Iglesias, Eduardo R. [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico)]. E-mail: iglesias@iie.org.mx; Flores Armenta, Magaly; Quijano Leon, Jose Luis; Torres Rodriguez, Marco A. [Comision Federal de Electricidad, Gerencia de Proyectos Geotermoelectricos, Morelia, Michoacan (Mexico); Torres, Rodolfo J.; Reyes Picasso, Neftali [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico); Sandoval Medina, Fernando [Comision Federal de Electricidad, Gerencia de Proyectos Geotermoelectricos, Morelia, Michoacan (Mexico)

2009-01-15

A previous tracer study in the Maritaro-La Cumbre area of the Los Azufres, Mich., geothermal field concluded that injection in well Az-15 recharges a group of producing wells located east of the injector to a maximum distance of about 2000 m. The study also indicated injection in Az-15 either does not recharge wells farther east in the neighborhood of the Laguna Verde fault, or does so at very low rates over a long period. Because it is important to recharge this zone, it was decided to explore the possibility of doing it by injecting in well Az-64, located in the area. Thus the present study was designed. Six producing wells were chosen to monitor tracer arrivals. Because all the chosen wells produce steam (several also produce water) and for economic reasons, it was decided to use a steam-phase tracer. We chose sulfur hexafluoride (SF6), a tracer successfully used before in the field. On October 24, 2006, we injected 96.4 kg of SF6 into well Az-64. All six chosen wells were sampled for 156 days, starting the day of injection. The tracer was detected in all six wells and for every well the shape of the residence curve clearly depicted the fractured nature of the corresponding permeability. Again in every case, the recovery curves indicated recovery was completed within the 156-day sampling period. The largest recoveries were in wells Az-05 (48.5%) and Az-43 (1.27%), those closest to the injection well and on the same side of the Maritaro fault as Az-64. Well Az-51, lying at 848 m from the injector and farthest away, registered the third-highest recovery percentage (1.01 x 10-5%), significantly higher than well Az-19 (6.95 x 10-6%), located 490 m from the injector; the two wells are on the same side of the Maritaro fault as Az-64. Well Az-32, located 423 m from the injector, but on the other side of the Maritaro fault, recovered 5.6 x 10-7% of the injected tracer, comparable to 8.67 x 10-7%, the amount recovered in well Az-13 on the same side of the Maritaro fault

AZ-101 Mixer Pump Demonstration Data Acquisition System and Gamma Cart Data Acquisition Control System Software Configuration Management Plan

International Nuclear Information System (INIS)

WHITE, D.A.

1999-01-01

This Software Configuration Management Plan (SCMP) provides the instructions for change control of the AZ1101 Mixer Pump Demonstration Data Acquisition System (DAS) and the Sludge Mobilization Cart (Gamma Cart) Data Acquisition and Control System (DACS)

Results of gas monitoring of double-shell flammable gas watch list tanks

International Nuclear Information System (INIS)

Wilkins, N.E.

1995-01-01

Tanks 103-SY; 101-AW; 103-, 104-, and 105-AN are on the Flammable Gas Watch List. Recently, standard hydrogen monitoring system (SHMS) cabinets have been installed in the vent header of each of these tanks. Grab samples have been taken once per week, and a gas chromatograph was installed on tank 104-AN as a field test. The data that have been collected since gas monitoring began on these tanks are summarized in this document

241-SY-101 DACS instrument problem screen (SCR 448) acceptance test procedure

International Nuclear Information System (INIS)

ERMI, A.M.

1999-01-01

The operability of the 241-SY-101 Data Acquisition and Control System (DACS) to provide proper control and monitoring of the mitigation mixer pump and instrumentation installed in the 241-SY-101 underground storage tank utilizing the [INSTPROB] screen will be systematically evaluated by the performance of this procedure

Descriptions and diagrams of the primary and annulus ventilation systems of the double-shell tank farms as of January 1988

International Nuclear Information System (INIS)

Blackman, A.E.; Waters, E.D.

1994-01-01

This document is a compilation of information describing the ventilation systems of the Double-Shell Tank farms (214-AN, -AP, -AW, -AW, -AY, -AZ, and -SY). A general description of the primary tank and annulus ventilation systems is given along with specific information on the high efficiency particulate air (HEPA) filters, condensers, preheaters, exhaust fans, and piping. This information is considered to be current as of January 1988. 38 refs, 20 figs, 30 tabs

Work plan for transition of SY-101 hydrogen mitigation test project data acquisition and control system (DACS-1)

International Nuclear Information System (INIS)

McClees, J.; Truitt, R.W.

1994-01-01

The purpose of this effort is to transfer operating and maintenance responsibility for the 241-SY-101 data acquisition and control system (DACS-1) from Los Alamos National Laboratory to Westinghouse Hanford Company. This work plan defines the tasks required for a successful turnover. It identifies DACS-1 transition, deliverables, responsible organizations and individuals, interfaces, cost, and schedule. The transition plan will discuss all required hardware, software, documentation, maintenance, operations, and training for use at Hanford Waste Tank 241-SY-101. The transfer of responsibilities for DACS-1 to WHC is contingent on final approval of applicable Acceptance for Beneficial Use documentation by Waste Tank Operations. The DACS-1 was designed to provide data monitoring, display, and storage for Tank 241-SY-101. The DACS-1 also provides alarm and control of all the hydrogen mitigation testing systems, as well as ancillary systems and equipment (HVAC, UPS, etc.) required to achieve safe and reliable operation of the testing systems in the tank

Flammable gas tank safety program: Technical basis for gas analysis and monitoring

International Nuclear Information System (INIS)

Sherwood, D.J.

1995-01-01

Flammable gases generated in radioactive liquids. Twenty-five high level radioactive liquid waste storage tanks located underground at the Hanford Site are on a Flammable Gas Watch List because they contain waste which tends to retain the gases generated in it until rather large quantities are available for sudden release to the tank head space; if a tank is full it has little dome space, and a flammable concentration of gases could be produced--even if the tank is ventilated. If the waste has no tendency to retain gas generated in it then a continual flammable gas concentration in the tank dome space is established by the gas production rate and the tank ventilation rate (or breathing rate for unventilated tanks); this is also a potential problem for Flammable Gas Watch List tanks, and perhaps other Hanford tanks too. All Flammable Gas Watch List tanks will be fitted with Standard Hydorgen Monitoring Systems so that their behavior can be observed. In some cases, such as tank 241-SY-101, the data gathered from such observations will indicate that tank conditions need to be mitigated so that gas release events are either eliminated or rendered harmless. For example, a mixer pump was installed in tank 241-SY-101; operating the pump stirs the waste, replacing the large gas release events with small releases of gas that are kept below twenty-five percent of the lower flammability limit by the ventilation system. The concentration of hydrogen measured in Hanford waste tanks is greater than that of any other flammable gas. Hydrogen levels measured with a Standard Hydrogen Monitoring System in excess of 0.6 volume percent will cause Westinghouse Hanford Company to consider actions which will decrease the amount of flammable gas in the tank

Mitigation of the most hazardous tank at the Hanford Site

International Nuclear Information System (INIS)

Reynolds, D.A.

1994-09-01

Various tanks at the Hanford Site have been declared to be unresolved safety problems. This means that the tank has the potential to be beyond the limits covered by the current safety documentation. Tank 241-SY-101 poses the greatest hazard. The waste stored in this tank has periodically released hydrogen gas which exceeds the lower flammable limits. A mixer pump was installed in this tank to stir the waste. Stirring the waste would allow the hydrogen to be released slowly in a controlled manner and mitigate the hazard associated with this tank. The testing of this mixer pump is reported in this document. The mixer pump has been successful in controlling the hydrogen concentration in the tank dome to below the flammable limit which has mitigated the hazardous gas releases

Acceptance test procedure, 241-SY-101 Flexible Receiver System, Phase III testing

International Nuclear Information System (INIS)

Ritter, G.A.

1994-01-01

This Acceptance Test Procedure is for the 241-SY-101 Flexible Receiver System, Phase III Testing. This procedure will test the sealing integrity of the Flexible Receiver System to ensure that release of waste and aerosols will be minimized during the removal of the test mixer pump from tank SY-101

Parametric Analyses of Heat Removal from High-Level Waste Tanks

International Nuclear Information System (INIS)

TRUITT, J.B.

2000-01-01

The general thermal hydraulics program GOTH-SNF was used to predict the thermal response of the waste in tanks 241-AY-102 and 241-AZ-102 when mixed by two 300 horsepower mixer pumps. This mixing was defined in terms of a specific waste retrieval scenario. Both dome and annulus ventilation system flow are necessary to maintain the waste within temperature control limits during the mixing operation and later during the sludge-settling portion of the scenario are defined

A Discussion of SY-101 Crust Gas Retention and Release Mechanisms

International Nuclear Information System (INIS)

Mendoza, D.P.; Mahoney, L.A.; Gauglitz, P.A.; Rassat, S.D.; Caley, S.M.

1999-01-01

The flammable gas hazard in Hanford waste tanks was made an issue by the behavior of double-shell Tank (DST) 241-SY-101 (SY-101). Shortly after SY-101 was filled in 1980, the waste level began rising periodically, due to the generation and retention of gases within the slurry, and then suddenly dropping as the gases were released. An intensive study of the tank's behavior revealed that these episodic releases posed a safety hazard because the released gas was flammable, and, in some cases, the volume of gas released was sufficient to exceed the lower flammability limit (LFL) in the tank headspace (Alleinann et al. 1993). A mixer pump was installed in SY-101 in late 1993 to prevent gases from building up in the settled solids layer, and the large episodic gas releases have since ceased (Allemann et al. 1994; Stewart et al. 1994; Brewster et al. 1995). However, the surface level of SY-101 has been increasing since at least 1995, and in recent months the level growth has shown significant and unexpected acceleration. Based on a number of observations and measurements, including data from the void fraction instrument (VFI), we have concluded that the level growth is caused largely by increased gas retention in the floating crust. In September 1998, the crust contained between about 21 and 43% void based on VFI measurements (Stewart et al. 1998). Accordingly, it is important to understand the dominant mechanisms of gas retention, why the gas retention is increasing, and whether the accelerating level increase will continue, diminish or even reverse. It is expected that the retained gas in the crust is flammable, with hydrogen as a major constituent. This gas inventory would pose a flammable gas hazard if it were to release suddenly. In May 1997, the mechanisms of bubble retention and release from crust material were the subject of a workshop. The evaluation of the crust and potential hazards assumed a more typical void of roughly 15% gas. It could be similar to

A Discussion of SY-101 Crust Gas Retention and Release Mechanisms

Energy Technology Data Exchange (ETDEWEB)

SD Rassat; PA Gauglitz; SM Caley; LA Mahoney; DP Mendoza

1999-02-23

The flammable gas hazard in Hanford waste tanks was made an issue by the behavior of double-shell Tank (DST) 241-SY-101 (SY-101). Shortly after SY-101 was filled in 1980, the waste level began rising periodically, due to the generation and retention of gases within the slurry, and then suddenly dropping as the gases were released. An intensive study of the tank's behavior revealed that these episodic releases posed a safety hazard because the released gas was flammable, and, in some cases, the volume of gas released was sufficient to exceed the lower flammability limit (LFL) in the tank headspace (Allemann et al. 1993). A mixer pump was installed in SY-101 in late 1993 to prevent gases from building up in the settled solids layer, and the large episodic gas releases have since ceased (Allemann et al. 1994; Stewart et al. 1994; Brewster et al. 1995). However, the surface level of SY-101 has been increasing since at least 1995, and in recent months the level growth has shown significant and unexpected acceleration. Based on a number of observations and measurements, including data from the void fraction instrument (VFI), we have concluded that the level growth is caused largely by increased gas retention in the floating crust. In September 1998, the crust contained between about 21 and 43% void based on VFI measurements (Stewart et al. 1998). Accordingly, it is important to understand the dominant mechanisms of gas retention, why the gas retention is increasing, and whether the accelerating level increase will continue, diminish or even reverse. It is expected that the retained gas in the crust is flammable, with hydrogen as a major constituent. This gas inventory would pose a flammable gas hazard if it were to release suddenly. In May 1997, the mechanisms of bubble retention and release from crust material were the subject of a workshop. The evaluation of the crust and potential hazards assumed a more typical void of roughly 15% gas. It could be similar to

Simulation and characterization of a Hanford high-level waste slurry

International Nuclear Information System (INIS)

Russell, R.L.; Smith, H.D.

1996-09-01

The baseline waste used for this simulant is a blend of wastes from tanks 101-AZ, 102-AZ, 106-C, and 102-AY that have been through water washing. However, the simulant used in this study represents a combination of tank waste slurries and should be viewed as an example of the slurries that might be produced by blending waste from various tanks. It does not imply that this is representative of the actual waste that will be delivered to the privatization contractor(s). This blended waste sludge simulant was analyzed for grain size distribution, theological properties both as a function of concentration and aging, and calcining characteristics. The grain size distribution allows a comparison with actual waste with respect to theological properties. Slurries with similar grain size distributions of the same phases are expected to exhibit similar theological properties. Rheological properties may also change because of changes in the slurry's particulate supernate chemistry due to aging. Low temperature calcination allows the potential for hazardous gas generation to be investigated

Mechanistic analysis of double-shell tank gas release

Energy Technology Data Exchange (ETDEWEB)

Allemann, R.T.; Antoniak, Z.I.; Friley, J.R.; Haines, C.E.; Liljegren, L.M.; Somasundaram, S.

1991-12-01

Pacific Northwest Laboratory (PNL) is studying possible mechanisms and fluid dynamics contributing to the periodic release of gases from the double-shell waste storage tanks at Hanford. This study is being conducted for Westinghouse Hanford Company (WHC), a contractor for the US Department of Energy (DOE). This interim report discusses the work done through November 1990. Safe management of the wastes at Hanford depends on an understanding of the chemical and physical mechanisms that take place in the waste tanks. An example of the need to understand these mechanisms is tank 101-SY. The waste in this tank is generating and periodically releasing potentially flammable gases into the tank vent system according to observations of the tank. How these gases are generated and become trapped, the causes of periodic release, and the mechanism of the release are not known in detail. In order to develop a safe mitigation strategy, possible physical mechanisms for the periodic release of flammable gases need to be understood.

Computer system requirements specification for 101-SY hydrogen mitigation test project data acquisition and control system (DACS-1)

International Nuclear Information System (INIS)

McNeece, S.G.; Truitt, R.W.

1994-01-01

The system requirements specification for SY-101 hydrogen mitigation test project (HMTP) data acquisition and control system (DACS-1) documents the system requirements for the DACS-1 project. The purpose of the DACS is to provide data acquisition and control capabilities for the hydrogen mitigation testing of Tank SY-101. Mitigation testing uses a pump immersed in the waste, directed at varying angles and operated at different speeds and time durations. Tank and supporting instrumentation is brought into the DACS to monitor the status of the tank and to provide information on the effectiveness of the mitigation test. Instrumentation is also provided for closed loop control of the pump operation. DACS is also capable for being expanded to control and monitor other mitigation testing. The intended audience for the computer system requirements specification includes the SY-101 hydrogen mitigation test data acquisition and control system designers: analysts, programmers, instrument engineers, operators, maintainers. It is intended for the data users: tank farm operations, mitigation test engineers, the Test Review Group (TRG), data management support staff, data analysis, Hanford data stewards, and external reviewers

Ion exchange removal of cesium from simulated and actual supernate from Hanford tanks 241-SY-101 and 241-SY-103

International Nuclear Information System (INIS)

Brown, G.N.; Bontha, J.R.; Carlson, C.D.

1995-09-01

Pacific Northwest Laboratory (PNL), in conjunction with the Process Chemistry and Statistics Section of Westinghouse Hanford Company (WHC), conducted this study as part of the Supernatant Treatment Development Task for the Initial Pretreatment Module (IPM) Applied Engineering Project. The study assesses the performance of the CS-100 ion exchange material for removing cesium from simulated and actual alkaline supernate from Hanford tanks 241-SY-101 and 241-SY-103. The objective of these experiments is to compare the cesium ion exchange loading and elution profiles of actual and simulated wastes. Specific experimental objectives include (1) demonstration of decontamination factors (DF) for cesium removal, 92) verification of simulant performance, (3) investigation of waste/exchanger chemistry, and (4) determination of the radionuclide content of the regenerated CS-100 resin prior to disposal

Advanced waste form and Melter development for treatment of troublesome high-level wastes

Energy Technology Data Exchange (ETDEWEB)

Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-10-01

A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe₂O₃ (also with high Al₂O₃ concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both

Design analysis supporting 101-SY Water Decon System

International Nuclear Information System (INIS)

Cleveland, K.J.

1995-01-01

This document contains the results of stress analysis and component sizing for the 101-SY mitigation pump, Water Decon System. Calculations included are a stress analysis of the High Pressure Manifold, the threaded connection on the Yoke Water Connector and a sizing of an air receiver tank

Technetium in alkaline, high-salt, radioactive tank waste supernate: Preliminary characterization and removal

International Nuclear Information System (INIS)

Blanchard, D.L. Jr.; Brown, G.N.; Conradson, S.D.

1997-01-01

This report describes the initial work conducted at Pacific Northwest National Laboratory to study technetium (Tc) removal from Hanford tank waste supernates and Tc oxidation state in the supernates. Filtered supernate samples from four tanks were studied: a composite double shell slurry feed (DSSF) consisting of 70% from Tank AW-101, 20% from AP-106, and 10% from AP-102; and three complexant concentrate (CC) wastes (Tanks AN-107, SY-101, ANS SY-103) that are distinguished by having a high concentration of organic complexants. The work included batch contacts of these waste samples with Reillex trademark-HPQ (anion exchanger from Reilly Industries) and ABEC 5000 (a sorbent from Eichrom Industries), materials designed to effectively remove Tc as pertechnetate from tank wastes. A short study of Tc analysis methods was completed. A preliminary identification of the oxidation state of non-pertechnetate species in the supernates was made by analyzing the technetium x-ray absorption spectra of four CC waste samples. Molybdenum (Mo) and rhenium (Re) spiked test solutions and simulants were tested with electrospray ionization-mass spectrometry to evaluate the feasibility of the technique for identifying Tc species in waste samples

Risk assessment methodology for Hanford high-level waste tanks

International Nuclear Information System (INIS)

Bott, T.F.; Mac Farlane, D.R.; Stack, D.W.; Kindinger, J.

1992-01-01

A methodology is presented for applying Probabilistic Safety Assessment techniques to quantification of the health risks posed by the high-level waste (HLW) underground tanks at the Department of Energy's Hanford reservation. This methodology includes hazard screening development of a list of potential accident initiators, systems fault trees development and quantification, definition of source terms for various release categories, and estimation of health consequences from the releases. Both airborne and liquid pathway releases to the environment, arising from aerosol and spill/leak releases from the tanks, are included in the release categories. The proposed methodology is intended to be applied to a representative subset of the total of 177 tanks, thereby providing a baseline risk profile for the HLW tank farm that can be used for setting clean-up/remediation priorities. Some preliminary results are presented for Tank 101-SY

241-SY-101 data acquisition and control system (DACS) remote operator interface operational test report

International Nuclear Information System (INIS)

ERMI, A.M.

1999-01-01

The readiness of the upgraded 241-SY-101 Data Acquisition and Control System (DACS) to provide proper control and monitoring of the mixer pump and instrumentation in tank 241-SY-101 was evaluated by the performance of OTP-440-001. Results of the OTP are reported here

Hanford Single-Shell Tank Leak Causes and Locations - 241-C Farm

Energy Technology Data Exchange (ETDEWEB)

Girardot, Crystal L.; Harlow, Donald G.

2013-07-30

This document identifies 241-C Tank Farm (C Farm) leak causes and locations for the 100 series leaking tanks (241-C-101 and 241-C-105) identified in RPP-RPT-33418, Rev. 2, Hanford C-Farm Leak Inventory Assessments Report. This document satisfies the C Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

Project Management Plan for Initial Tank Retrieval Systems, Project W-211

International Nuclear Information System (INIS)

VAN BEEK, J.E.

1999-01-01

Project W-211, Initial Tank Retrieval Systems (ITRS), is a fiscal year 1994 Major Systems Acquisition that will provide systems for retrieval of radioactive wastes from selected double-shell tanks (DST). The contents of these tanks are a combination of supernatant liquids and settled solids. To retrieve waste from the tanks, it is first necessary to mix the liquid and solids prior to transferring the slurry to alternative storage or treatment facilities. The ITRS will provide systems to mobilize the settled solids and transfer the wastes out of the tanks. In so doing, ITRS provides feed for future processing plants, allows for consolidation of tank solids to manage space within existing DST storage capacity, and supports continued safe storage of tank waste. The ITRS scope has been revised to include waste retrieval systems for tanks AP-102, AP-104, AP-108, AN-103, AN-104, AN-105, AY-102, AZ-102, and SY-102. This current tank selection and sequence provides retrieval systems supporting the Privatized waste processing plant and sustains the ability to provide final remediation of several watch list DSTs via treatment. The ITRS is configured to support changing program needs, as constrained by available budget, by maintaining the flexibility for exchanging tanks requiring mixer pump-based retrieval systems and shifting the retrieval sequence. Preliminary design was configured such that an adequate basis exists for initiating Title II design of a mixer pump based retrieval system for any DST. This Project Management Plan (PMP) documents the methodology for managing the ITRS, formalizes organizational responsibilities and interfaces, and identifies project requirements such as change control, design verification, systems engineering, and human factors engineering

Hazard assessments of double-shell flammable gas tanks

International Nuclear Information System (INIS)

Fox, G.L.; Stepnewski, D.D.

1994-01-01

This report is the fourth in a series of hazard assessments performed on the double-shell flammable gas watch list tanks. This report focuses on hazards associated with the double-shell watch list tanks (101-AW, 103-AN, 104-AN, and 105-AN). While a similar assessment has already been performed for tank 103-SY, it is also included here to incorporate a more representative slurry gas mixture and provide a consistent basis for comparing results for all the flammable gas tanks. This report is intended to provide an in-depth assessment by considering the details of the gas release event and slurry gas mixing as the gas is released from the waste. The consequences of postulated gas ignition are evaluated using a plume burn model and updated ignition frequency predictions. Tank pressurization which results from a gas burn, along with the structural response, is also considered. The report is intended to support the safety basis for work activities in flammable gas tanks by showing margins to safety limits that are available in the design and procedures

Hanford double shell tank corrosion monitoring instrument tree prototype

International Nuclear Information System (INIS)

Nelson, J.L.; Edgemon, G.L.; Ohl, P.C.

1995-11-01

High-level nuclear wastes at the Hanford site are stored underground in carbon steel double-shell and single-shell tanks (DSTs and SSTs). The installation of a prototype corrosion monitoring instrument tree into DST 241-A-101 was completed in December 1995. The instrument tree has the ability to detect and discriminate between uniform corrosion, pitting, and stress corrosion cracking (SCC) through the use of electrochemical noise measurements and a unique stressed element, three-electrode probe. The tree itself is constructed of AISI 304L stainless steel (UNS S30403), with probes in the vapor space, vapor/liquid interface and liquid. Successful development of these trees will allow their application to single shell tanks and the transfer of technology to other US Department of Energy (DOE) sites. Keywords: Hanford, radioactive waste, high-level waste tanks, electrochemical noise, probes, double-shell tanks, single-shell tanks, corrosion

Characterization of mussel H2A.Z.2: a new H2A.Z variant preferentially expressed in germinal tissues from Mytilus.

Science.gov (United States)

Rivera-Casas, Ciro; González-Romero, Rodrigo; Vizoso-Vazquez, Ángel; Cheema, Manjinder S; Cerdán, M Esperanza; Méndez, Josefina; Ausió, Juan; Eirin-Lopez, Jose M

2016-10-01

Histones are the fundamental constituents of the eukaryotic chromatin, facilitating the physical organization of DNA in chromosomes and participating in the regulation of its metabolism. The H2A family displays the largest number of variants among core histones, including the renowned H2A.X, macroH2A, H2A.B (Bbd), and H2A.Z. This latter variant is especially interesting because of its regulatory role and its differentiation into 2 functionally divergent variants (H2A.Z.1 and H2A.Z.2), further specializing the structure and function of vertebrate chromatin. In the present work we describe, for the first time, the presence of a second H2A.Z variant (H2A.Z.2) in the genome of a non-vertebrate animal, the mussel Mytilus. The molecular and evolutionary characterization of mussel H2A.Z.1 and H2A.Z.2 histones is consistent with their functional specialization, supported on sequence divergence at promoter and coding regions as well as on varying gene expression patterns. More precisely, the expression of H2A.Z.2 transcripts in gonadal tissue and its potential upregulation in response to genotoxic stress might be mirroring the specialization of this variant in DNA repair. Overall, the findings presented in this work complement recent reports describing the widespread presence of other histone variants across eukaryotes, supporting an ancestral origin and conserved role for histone variants in chromatin.

Oxidative-Alkaline Leaching of Washed 241-SY-102 and 241-SX-101 Tank Sludges and Its Impact on Immobilized High-Level Waste

International Nuclear Information System (INIS)

Rapko, Brian M.; Geeting, John GH; Sinkov, Sergei I.; Vienna, John D.

2006-01-01

This report describes work designed to evaluate the effectiveness of alkaline permanganate contacts at selectively removing chromium from the Hanford tank sludges 241-SY-102 and 241-SX-101. The key variables examined in this study, as compared to contact with the standard conditions of stoichiometric permanganate in 3 M hydroxide at elevated temperature, were: (a) excess permanganate and hydroxide at elevated temperature, (b) the separation of an elevated temperature 3 M hydroxide leach with either a room temperature permanganate contact or an elevated temperature permanganate contact at 0.25 M hydroxide. It was determined that sequential permanganate and caustic leaching can provide as effective removal of Cr as the combined high hydroxide permanganate contact at elevated temperature while minimizing concomitant Pu dissolution

Double-Shell Tank (DST) Ventilation System Vapor Sampling and Analysis Plan

International Nuclear Information System (INIS)

SASAKI, L.M.

2000-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for vapor samples from the primary ventilation systems of the AN, AP, AW, and AY/AZ tank farms. Sampling will be performed in accordance with Data Quality Objectives for Regulatory Requirements for Hazardous and Radioactive Air Emissions Sampling and Analysis (Air DQO) (Mulkey 1999). The sampling will verify if current air emission estimates used in the permit application are correct and provide information for future air permit applications. Vapor samples will be obtained from tank farm ventilation systems, downstream from the tanks and upstream of any filtration. Samples taken in support of the DQO will consist of SUMMA(trademark) canisters, triple sorbent traps (TSTs), sorbent tube trains (STTs), polyurethane foam (PUF) samples. Particulate filter samples and tritium traps will be taken for radiation screening to allow the release of the samples for analysis. The following sections provide the general methodology and procedures to be used in the preparation, retrieval, transport, analysis, and reporting of results from the vapor samples

Physical mechanisms contributing to the episodic gas release from Hanford tank 241-SY-101

International Nuclear Information System (INIS)

Allemann, R.T.

1992-04-01

Volume growth of contents in a waste storage tank at Hanford is accompanied by episodic releases of gas and a rise in the level of tank contents. A theory is presented to describe how the gas is retained in the waste and how it is released. The theory postulates that somewhat cohesive gobs of sludge rise from the lower regions of the tank and buoyancy overcomes the cohesive strength of the slurry; this quantitatively explains several of the measured phenomena and qualitatively explains other observations

Test procedures and instructions for Hanford tank waste supernatant cesium removal

Energy Technology Data Exchange (ETDEWEB)

Hendrickson, D.W., Westinghouse Hanford

1996-05-31

This document provides specific test procedures and instructions to implement the test plan for the preparation and conduct of a cesium removal test using Hanford Double-Shell Slurry Feed supernatant liquor from tank 251-AW-101 in a bench-scale column.Cesium sorbents to be tested include resorcinol-formaldehyde resin and crystalline silicotitanate. The test plan for which this provides instructions is WHC-SD-RE-TP-022, Hanford Tank Waste Supernatant Cesium Removal Test Plan.

Flammable Gas Safety Program: analysis of gas sampling probe locations in the SX-farm flammable gas watchlist tanks

International Nuclear Information System (INIS)

McLaren, J.M.; Claybrook, S.W.

1995-09-01

An analysis was performed to determine the optimum ventilation line up for the AN Tank Farm. The analysis used the postulated maximum historical GRE in tanks AN-103, -104, and -105. Tank AN-104 was found to be limiting. The results of the analysis show that an airflow of 250 cfm through tanks 241-AN-103, -104, and -105 with an airflow of 100 cfm through tanks 241-AN-101, -102, -106, and -107 would be the optimum ventilation lineup

Effect of current frequency on the mechanical properties, microstructure and texture evolution in AZ31 magnesium alloy strips during electroplastic rolling

Energy Technology Data Exchange (ETDEWEB)

Li, Xiaopei [Advanced Materials Institute, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055 (China); Beijing National Center for Electron Microscopy, School of Materials Science and Engineering, The State Key Laboratory of New Ceramics and Fine Processing, Laboratory of Advanced Materials (MOE), Tsinghua University, Beijing 100084 (China); Tang, Guoyi; Kuang, Jie; Li, Xiaohui [Advanced Materials Institute, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055 (China); Zhu, Jing, E-mail: jzhu@tsinghua.edu.cn [Advanced Materials Institute, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055 (China); Beijing National Center for Electron Microscopy, School of Materials Science and Engineering, The State Key Laboratory of New Ceramics and Fine Processing, Laboratory of Advanced Materials (MOE), Tsinghua University, Beijing 100084 (China)

2014-08-26

The effect of electroplastic rolling (ER) on the mechanical properties, microstructure and texture in the AZ31 magnesium alloy strips has been investigated by tensile testing and electron back scattered diffraction (EBSD) methods. It is shown that the mechanical properties, microstructure, and texture are highly current frequency-dependent. Best mechanical properties are obtained from the 500 Hz ER specimen by carrying out tensile tests for all the rolled strips. Besides, the frequencies of twin boundaries, which are reduced to the minimum at 500 Hz, vary with the current frequency. Moreover, it can be seen from the calculated (0001) and (101{sup ¯}0) pole figures that texture evolved into an obvious off-basal texture, and non-basal slip systems are activated under 500 Hz. The mechanisms of twinning growth and texture evolution in AZ31 magnesium alloy strips during ER are considered to be responsible for the experimental results.

Corrosion product layers on magnesium alloys AZ31 and AZ61: Surface chemistry and protective ability

Science.gov (United States)

Feliu, S.; Llorente, I.

2015-08-01

This paper studies the chemical composition of the corrosion product layers formed on magnesium alloys AZ31 and AZ61 following immersion in 0.6 M NaCl, with a view to better understanding their protective action. Relative differences in the chemical nature of the layers were quantified by X-ray photoelectron spectroscopy (XPS), scanning electron microscopy (SEM), energy dispersive analysis of X-ray (EDX) and low-angle X-ray diffraction (XRD). Corrosion behavior was investigated by Electrochemical Impedance Spectroscopy (EIS) and hydrogen evolution measurement. An inhibitive effect from the corrosion product layers was observed from EIS, principally in the case of AZ31, as confirmed by hydrogen evolution tests. A link was found between carbonate enrichment observed by XPS in the surface of the corrosion product layer, concomitant with the increase in the protective properties observed by EIS.

TWRS privatization support project waste characterization database development. Volume 1

International Nuclear Information System (INIS)

Brevick, C.H.

1995-11-01

Pacific Northwest National Laboratory requested support from ICF Kaiser Hanford Company in assembling radionuclide and chemical analyte sample data and inventory estimates for fourteen Hanford under-ground storage tanks: 241-AN-102, -104, -105, -106, and -107, 241-AP-102, -104, and -105; 241-AW-101, -103, and -105, 241-AZ-101 and-102; and 241-C-109. Sample data were assembled for sixteen radio nuclides and thirty five chemical analytes. The characterization data were provided to Pacific Northwest National Laboratory in support of the Tank Waste Remediation Services Privatization Support Project. The purpose of this report is to present the results and document the methodology used in preparing the waste characterization information data set to support the Tank Waste Remediation Services Privatization Support Project. This report describes the methodology used in assembling the waste characterization information and how that information was validated by a panel of independent technical reviewers. Also, contained in this report are the various data sets created., the master data set, a subset, and an unreviewed data set

241-SY-101 data acquisition and control system (DACS) operator interface upgrade operational test report

International Nuclear Information System (INIS)

ERMI, A.M.

1999-01-01

This procedure provides instructions for readiness of the first portion of the upgraded 241-SY-101 Data Acquisition and Control System (DACS) computer system to provide proper control and monitoring of the mitigation mixer pump and instrumentation installed in the 241-SY-101 underground storage tank will be systematically evaluated by the performance of this procedure

Evaluation of pressure transducers to measure surface level in the waste storage tanks

International Nuclear Information System (INIS)

Peters, T.J.; Colson, J.B.

1994-06-01

This report describes the results of tests conducted at the Pacific Northwest Laboratory (PNL) to determine if pressure transducers can be used to measure the surface level in the waste tanks. A survey was first conducted to evaluate which, if any, commercially available pressure transducers were available that could meet the requirements for use in the waste tanks. More than 35 companies were contacted to determine if they manufactured a pressure transducer that could be used in the 101-SY waste tank. The three basic requirements for a pressure transducer for this application were that they were radiation-hardened, could withstand a caustic environment, and were certified to be intrinsically safe. No manufacturer was able to meet all three of these requirements with a commercially available product. Seven companies were able to meet the requirements for being radiation-hardened and being able to withstand the caustic environment. However, only two of the nine companies were willing to supply a pressure transducer for laboratory testing. The two pressure transducers that were tested in this program were the VEGA D36-38 from HiTech Technologies, Inc., and the KP-1911-A from Kaman Instrumentation Corporation. Pressure transducers operate on the principle that the pressure at the location of a sensor increases directly with the depth of the liquid above it. A liquid is required in order for these devices to operate. For these tests, water was first used to determine the ideal operation of the devices, then the devices were placed in a 101-SY waste tank simulant. The simulant had a specific gravity of 1.96 and had the consistency similar to the convective layer in the 101-SY waste tank. In order to determine the surface level with pressure transducers, the density of the material needs to be known

ICPP tank farm closure study. Volume 1

International Nuclear Information System (INIS)

Spaulding, B.C.; Gavalya, R.A.; Dahlmeir, M.M.

1998-02-01

The disposition of INEEL radioactive wastes is now under a Settlement Agreement between the DOE and the State of Idaho. The Settlement Agreement requires that existing liquid sodium bearing waste (SBW), and other liquid waste inventories be treated by December 31, 2012. This agreement also requires that all HLW, including calcined waste, be disposed or made road ready to ship from the INEEL by 2035. Sodium bearing waste (SBW) is produced from decontamination operations and HLW from reprocessing of SNF. SBW and HLW are radioactive and hazardous mixed waste; the radioactive constituents are regulated by DOE and the hazardous constituents are regulated by the Resource Conservation and Recovery Act (RCRA). Calcined waste, a dry granular material, is produced in the New Waste Calcining Facility (NWCF). Two primary waste tank storage locations exist at the ICPP: Tank Farm Facility (TFF) and the Calcined Solids Storage Facility (CSSF). The TFF has the following underground storage tanks: four 18,400-gallon tanks (WM 100-102, WL 101); four 30,000-gallon tanks (WM 103-106); and eleven 300,000+ gallon tanks. This includes nine 300,000-gallon tanks (WM 182-190) and two 318,000 gallon tanks (WM 180-181). This study analyzes the closure and subsequent use of the eleven 300,000+ gallon tanks. The 18,400 and 30,000-gallon tanks were not included in the work scope and will be closed as a separate activity. This study was conducted to support the HLW Environmental Impact Statement (EIS) waste separations options and addresses closure of the 300,000-gallon liquid waste storage tanks and subsequent tank void uses. A figure provides a diagram estimating how the TFF could be used as part of the separations options. Other possible TFF uses are also discussed in this study

ICPP tank farm closure study. Volume 1

Energy Technology Data Exchange (ETDEWEB)

Spaulding, B.C.; Gavalya, R.A.; Dahlmeir, M.M. [and others

1998-02-01

The disposition of INEEL radioactive wastes is now under a Settlement Agreement between the DOE and the State of Idaho. The Settlement Agreement requires that existing liquid sodium bearing waste (SBW), and other liquid waste inventories be treated by December 31, 2012. This agreement also requires that all HLW, including calcined waste, be disposed or made road ready to ship from the INEEL by 2035. Sodium bearing waste (SBW) is produced from decontamination operations and HLW from reprocessing of SNF. SBW and HLW are radioactive and hazardous mixed waste; the radioactive constituents are regulated by DOE and the hazardous constituents are regulated by the Resource Conservation and Recovery Act (RCRA). Calcined waste, a dry granular material, is produced in the New Waste Calcining Facility (NWCF). Two primary waste tank storage locations exist at the ICPP: Tank Farm Facility (TFF) and the Calcined Solids Storage Facility (CSSF). The TFF has the following underground storage tanks: four 18,400-gallon tanks (WM 100-102, WL 101); four 30,000-gallon tanks (WM 103-106); and eleven 300,000+ gallon tanks. This includes nine 300,000-gallon tanks (WM 182-190) and two 318,000 gallon tanks (WM 180-181). This study analyzes the closure and subsequent use of the eleven 300,000+ gallon tanks. The 18,400 and 30,000-gallon tanks were not included in the work scope and will be closed as a separate activity. This study was conducted to support the HLW Environmental Impact Statement (EIS) waste separations options and addresses closure of the 300,000-gallon liquid waste storage tanks and subsequent tank void uses. A figure provides a diagram estimating how the TFF could be used as part of the separations options. Other possible TFF uses are also discussed in this study.

Project Execution Plan for Project W-211 Initial Tank Retrieval Systems (ITRS)

International Nuclear Information System (INIS)

VAN BEEK, J.E.

2000-01-01

This Project Execution Plan documents the methodology for managing Project W-211. Project W-211, Initial Tank Retrieval Systems (ITRS), is a fiscal year 1994 Major Systems Acquisition that will provide systems for retrieval of radioactive wastes from selected double-shell tanks (DST). The contents of these tanks are a combination of supernatant liquids and settled solids. To retrieve waste from the tanks, it is first necessary to mix the liquid and solids prior to transferring the slurry to alternative storage or treatment facilities. The ITRS will provide systems to mobilize the settled solids and transfer the wastes out of the tanks. In so doing, ITRS provides feed for the future waste treatment plant, allows for consolidation of tank solids to manage space within existing DST storage capacity, and supports continued safe storage of tank waste. The ITRS scope has been revised to include waste retrieval systems for tanks AP-102, AP-104, AN-102, AN-103, AN-104, AN-105, AY-102, AZ-102, and SY-102. This current tank selection and sequence provides retrieval systems supporting the River Protection Project (RF'P) Waste Treatment Facility and sustains the ability to provide final remediation of several watch list DSTs via treatment. The ITRS is configured to support changing program needs, as constrained by available budget, by maintaining the flexibility for exchanging tanks requiring mixer pump-based retrieval systems and shifting the retrieval sequence. Preliminary design was configured such that an adequate basis exists for initiating Title II design of a mixer pump-based retrieval system for any DST. This Project Execution Plan (PEP), derived from the predecessor Project Management Plan, documents the methodology for managing the ITRS, formalizes organizational responsibilities and interfaces, and identifies project requirements such as change control, design verification, systems engineering, and human factors engineering

Washing and Caustic Leaching of Hanford Tank Sludge: Results of FY 1998 Studies

Energy Technology Data Exchange (ETDEWEB)

GJ Lumetta; BM Rapko; J Liu; DJ Temer; RD Hunt

1998-12-11

Sludge washing and parametric caustic leaching tests were performed on sludge samples tiom five Hanford tanks: B-101, BX-1 10, BX-112, C-102, and S-101. These studies examined the effects of both dilute hydroxide washing and caustic leaching on the composition of the residual sludge solids. ` Dilute hydroxide washing removed from <1 to 25% of the Al, -20 to 45% of the Cr, -25 to 97% of the P, and 63 to 99% of the Na from the Hdord tank sludge samples examined. The partial removal of these elements was likely due to the presence of water-soluble sodium salts of aluminate, chromate, hydroxide, nitrate, nitrite, and phosphate, either in the interstitial liquid or as dried salts.

77 FR 10649 - Modification of Class E Airspace; Douglas, AZ

Science.gov (United States)

2012-02-23

...-1313; Airspace Docket No. 11-AWP-17] Modification of Class E Airspace; Douglas, AZ AGENCY: Federal... Bisbee Douglas International Airport, Douglas, AZ. Decommissioning of the Cochise VHF Omni-Directional... Federal Register a notice of proposed rulemaking to amend controlled airspace at Douglas, AZ (76 FR 78180...

Multi-Function Waste Tank Facility thermal hydraulic analysis for Title II design

International Nuclear Information System (INIS)

Cramer, E.R.

1994-01-01

The purpose of this work was to provide the thermal hydraulic analysis for the Multi-Function Waste Tank Facility (MWTF) Title II design. Temperature distributions throughout the tank structure were calculated for subsequent use in the structural analysis and in the safety evaluation. Calculated temperatures of critical areas were compared to design allowables. Expected operating parameters were calculated for use in the ventilation system design and in the environmental impact documentation. The design requirements were obtained from the MWTF Functional Design Criteria (FDC). The most restrictive temperature limit given in the FDC is the 200 limit for the haunch and dome steel and concrete. The temperature limit for the rest of the primary and secondary tanks and concrete base mat and supporting pad is 250 F. Also, the waste should not be allowed to boil. The tank geometry was taken from ICF Kaiser Engineers Hanford drawing ES-W236A-Z1, Revision 1, included here in Appendix B. Heat removal rates by evaporation from the waste surface were obtained from experimental data. It is concluded that the MWTF tank cooling system will meet the design temperature limits for the design heat load of 700,000 Btu/h, even if cooling flow is lost to the annulus region, and temperatures change very slowly during transients due to the high heat capacity of the tank structure and the waste. Accordingly, transients will not be a significant operational problem from the viewpoint of meeting the specified temperature limits

Effects of Crust Ingestion on Mixer Pump Performance in Tank 241-SY-101: Workshop Results

International Nuclear Information System (INIS)

Brennen, C.E.; Stewart, C.W.; Meyer, P.A.

1999-01-01

In August 1999, a workshop was held at Pacific Northwest National Laboratory to discuss the effects of crust ingestion on mixer pump performance in Hanford Waste Tank 241-SY-101. The main purpose of the workshop was to evaluate the potential for crust ingestion to degrade mixing and/or damage the mixer pump. The need for a previously determined 12-inch separation between the top of the mixer pump inlet and the crust base was evaluated. Participants included a representative from the pump manufacturer, an internationally known expert in centrifugal pump theory, Hanford scientists and engineers, and operational specialists representing relevant fields of expertise. The workshop focused on developing an understanding of the pump design, addressing the physics of entrainment of solids and gases into the pump, and assessing the effects of solids and gases on pump performance. The major conclusions are summarized as follows: (1) Entrainment of a moderate amount of solids or gas from the crust should not damage the pump or reduce its lifetime, though mixing effectiveness will be somewhat reduced. (2) Air binding should not damage the pump. Vibrations due to ingestion of gas, solids, and objects potentially could cause radial loads that might reduce the lifetime of bearings and seals. However, significant damage would require extreme conditions not associated with the small bubbles, fine solids, and chunks of relatively weak material typical of the crust. (3) The inlet duct extension opening, 235 inches from the tank bottom, should be considered the pump inlet, not the small gap at 262 inches. (4) A suction vortex exists at the inlet of all pumps. The characteristics of the inlet suction vortex in the mixer pump are very hard to predict, but its effects likely extend upward several feet. Because of this, the current 12-inch limit should be replaced with criteria based on actual monitored pump performance. The most obvious criterion (in addition to current operational

241-SY-101 mixer pump lifetime expectancy. Final report

International Nuclear Information System (INIS)

Shaw, C.P.

1995-01-01

The purpose of WHC-SD-WM-TI-726, Rev. 0 241-SY-101 Mixer Pump Lifetime Expectancy is to determine a best estimate of the mean lifetime of non-repairable (located in the waste) essential features of the hydrogen mitigation mixer pump presently installed in 101-SY. The estimated mean lifetime is 9.1 years. This report does not demonstrate operation of the entire pump assembly within the Tank Farm ''safety envelope''. It was recognized by the Defense Nuclear Facilities Safety Board (DNFSB) this test pump was not specifically designed for long term service in tank 101-SY. In June 95 the DNFSB visited Hanford and ask the question, ''how long will this test pump last and how will the essential features fail?'' During the 2 day meeting with the DNFSB it was discussed and defined within the meeting just exactly what essential features of the pump must operate. These essential features would allow the pump to operate for the purpose of extending the window for replacement. Operating with only essential features would definitely be outside the operating safety envelope and would require a waiver. There are three essential features: 1. The pump itself (i.e. the impeller and motor) must operate 2. Nozzles and discharges leg must remain unplugged 3. The pump can be re-aimed, new waste targeted, even if manually

Contingency plan for deployment of the void fraction instrument in Tank 241-AY-102

International Nuclear Information System (INIS)

CONNER, J.M.

1999-01-01

High-heat producing sludge from tank 241-C-106 will be sluiced and transferred to tank 241-AY-102 beginning in October 1998. Safety analyses have postulated that after retrieval, the waste in 241-AY-102 may generate and retain unsafe levels of flammable gases (Noorani 1998, Pasamebmetoglu etal. 1997). Unsafe levels of retained gas are not expected, but cannot be ruled out because of the large uncertainty in the gas generation and retention rates. The Tank Waste Remediation System Basis for Interim Operation (Noorani 1998) identifies the need for a contingency plan to add void fraction monitoring to tank 241-AY-102 within 2 weeks of the identification of flammable gas buildup that would warrant monitoring. The Tank 241-C-106 Waste Retrieval Sluicing System Process Control Plan (Carothers et al. 1998) committed to providing a contingency plan for deployment of the void fraction instrument (VFI) in tank 241-AY-102. The VFI determines the local void fraction of the waste by compressing a waste sample captured in a gas-tight test chamber. The sample chamber is mounted on the end of a 76-cm (2.5-ft) arm that can be rotated from vertical to horizontal when the instrument is deployed. Once in the waste, the arm can be positioned horizontally and rotated to sample in different areas below the riser. The VFI is deployed using a crane. The VFI has been deployed previously in 241-AW, 241-AN, and 241-SY tank farms, most recently in tank 241-SY-101 in June and July 1998. An additional test in tank 241-SY-101 is planned in September 1998. Operating instructions for the VFI are included in the Void Fraction Instrument Operation and Maintenance Manual (Pearce 1994)

SY Tank Farm ventilation isolation option risk assessment report

Energy Technology Data Exchange (ETDEWEB)

Powers, T.B.; Morales, S.D.

1994-03-01

The safety of the 241-SY Tank Farm ventilation system has been under extensive scrutiny due to safety concerns associated with tank 101-SY. Hydrogen and other gases are generated and trapped in the waste below the liquid surface. Periodically, these gases are released into the dome space and vented through the exhaust system. This attention to the ventilation system has resulted in the development of several alternative ventilation system designs. The ventilation system provides the primary means of mitigation of accidents associated with flammable gases. This report provides an assessment of various alternatives ventilation system designs.

Calcination/dissolution testing for Hanford Site tank wastes

International Nuclear Information System (INIS)

Colby, S.A.; Delegard, C.H.; McLaughlin, D.F.; Danielson, M.J.

1994-07-01

Thermal treatment by calcination offers several benefits for the treatment of Hanford Site tank wastes, including the destruction of organics and ferrocyanides and an hydroxide fusion that permits the bulk of the mostly soluble nonradioactive constituents to be easily separated from the insoluble transuranic residue. Critical design parameters were tested, including: (1) calciner equipment design, (2) hydroxide fusion chemistry, and (3) equipment corrosion. A 2 gal/minute pilot plant processed a simulated Tank 101-SY waste and produced a free flowing 700 C molten calcine with an average calciner retention time of 20 minutes and >95% organic, nitrate, and nitrite destruction. Laboratory experiments using actual radioactive tank waste and the simulated waste pilot experiments indicate that 98 wt% of the calcine produced is soluble in water, leaving an insoluble transuranic fraction. All of the Hanford Site tank wastes can benefit from calcination/dissolution processing, contingent upon blending various tank waste types to ensure a target of 70 wt% sodium hydroxide/nitrate/nitrite fluxing agent. Finally, corrosion testing indicates that a jacketed nickel liner cooled to below 400 C would corrode <2 mil/year (0.05 mm/year) from molten calcine attack

Evaluating Residence Time for Cesium Removal from Simulated Hanford Tank Wastes Using SuperLig(R) 644 Resin

International Nuclear Information System (INIS)

Hassan, N.M.

2003-01-01

Batch contact and column experiments were performed to evaluate the effect of residence time on cesium (Cs) removal from two simulated Hanford tank wastes using SuperLig(R) 644 resin. The two waste simulants mimic the compositions of tanks 241-AZ-102 and 241-AN-107 at the U.S. Department of Energy (DOE) Hanford site. A single column made of glass tube (2.7-cm i.d.), which contained approximately 100 mL of H-form SuperLig(R) 644 resin was used in the column experiments. The experiments each consisted of loading, elution, and regeneration steps were performed at flow rates ranging from 0.64 to 8.2 BV/h for AZ-102 and from 1.5 to 18 BV/h for AN-107 simulant. The lowest flow rates of 0.64 and 1.5 BV/h were selected to evaluate less than optimal flow conditions in the plant. The range of the flow rates is consistent with the River Protection Project design for the waste treatment plant (WTP) columns, which will operate at a flow rate between 1.5 to 3 BV/h. Batch contact experiments were also performed for two batches of SuperLig(R) 644 to determine the equilibrium distribution coefficients (Kds) as a function of Cs concentration

Mechanistic analysis of double-shell tank gas release. Progress report, November 1990

Energy Technology Data Exchange (ETDEWEB)

Allemann, R.T.; Antoniak, Z.I.; Friley, J.R.; Haines, C.E.; Liljegren, L.M.; Somasundaram, S.

1991-12-01

Pacific Northwest Laboratory (PNL) is studying possible mechanisms and fluid dynamics contributing to the periodic release of gases from the double-shell waste storage tanks at Hanford. This study is being conducted for Westinghouse Hanford Company (WHC), a contractor for the US Department of Energy (DOE). This interim report discusses the work done through November 1990. Safe management of the wastes at Hanford depends on an understanding of the chemical and physical mechanisms that take place in the waste tanks. An example of the need to understand these mechanisms is tank 101-SY. The waste in this tank is generating and periodically releasing potentially flammable gases into the tank vent system according to observations of the tank. How these gases are generated and become trapped, the causes of periodic release, and the mechanism of the release are not known in detail. In order to develop a safe mitigation strategy, possible physical mechanisms for the periodic release of flammable gases need to be understood.

Results of Large-Scale Testing on Effects of Anti-Foam Agent on Gas Retention and Release

Energy Technology Data Exchange (ETDEWEB)

Stewart, Charles W.; Guzman-Leong, Consuelo E.; Arm, Stuart T.; Butcher, Mark G.; Golovich, Elizabeth C.; Jagoda, Lynette K.; Park, Walter R.; Slaugh, Ryan W.; Su, Yin-Fong; Wend, Christopher F.; Mahoney, Lenna A.; Alzheimer, James M.; Bailey, Jeffrey A.; Cooley, Scott K.; Hurley, David E.; Johnson, Christian D.; Reid, Larry D.; Smith, Harry D.; Wells, Beric E.; Yokuda, Satoru T.

2008-01-03

The U.S. Department of Energy (DOE) Office of River Protection’s Waste Treatment Plant (WTP) will process and treat radioactive waste that is stored in tanks at the Hanford Site. The waste treatment process in the pretreatment facility will mix both Newtonian and non-Newtonian slurries in large process tanks. Process vessels mixing non-Newtonian slurries will use pulse jet mixers (PJMs), air sparging, and recirculation pumps. An anti-foam agent (AFA) will be added to the process streams to prevent surface foaming, but may also increase gas holdup and retention within the slurry. The work described in this report addresses gas retention and release in simulants with AFA through testing and analytical studies. Gas holdup and release tests were conducted in a 1/4-scale replica of the lag storage vessel operated in the Pacific Northwest National Laboratory (PNNL) Applied Process Engineering Laboratory using a kaolin/bentonite clay and AZ-101 HLW chemical simulant with non-Newtonian rheological properties representative of actual waste slurries. Additional tests were performed in a small-scale mixing vessel in the PNNL Physical Sciences Building using liquids and slurries representing major components of typical WTP waste streams. Analytical studies were directed at discovering how the effect of AFA might depend on gas composition and predicting the effect of AFA on gas retention and release in the full-scale plant, including the effects of mass transfer to the sparge air. The work at PNNL was part of a larger program that included tests conducted at Savannah River National Laboratory (SRNL) that is being reported separately. SRNL conducted gas holdup tests in a small-scale mixing vessel using the AZ-101 high-level waste (HLW) chemical simulant to investigate the effects of different AFAs, their components, and of adding noble metals. Full-scale, single-sparger mass transfer tests were also conducted at SRNL in water and AZ-101 HLW simulant to provide data for PNNL�

Results of shear studies with 241-AY-101 sludge

International Nuclear Information System (INIS)

WARRANT, R.W.

2001-01-01

The Department of Energy's Tanks Focus Area (TFA) authorized a project to study the effect of shear on the settling properties of high-level waste sludge to support retrieval programs. A series of settling studies was conducted on a composite sample of tank 241-AY-101 (AY-101) material. Comparisons were made with duplicate samples that were sheared with a tissue homogenizer and allowed to settle. Aliquots of sheared and unsheared settled solids were submitted for chemical and radiological analyses. There are five major conclusions from the study that apply to AY-101 sludge: (1) Sludge settling rates are detectably decreased after shearing of particles by means of a tissue homogenizer. A significant decrease in the settling rates was measured after 2 minutes of shearing. A smaller additional decrease in the settling rates was observed after an additional 10 minutes of shearing. (2) Sodium and Cesium appear to be present in both the liquid and solid phases of the composite sample. (3) The shearing of the solids does not appear to significantly change the distribution of the radionuclides, ( 241 Am, 90 Sr, Total Alpha, or other radionuclides), within the solids. (4) The mean particle diameter decreases after shearing with the tissue homogenizer and affects the settling rate in proportion to the square of the particle diameter. (5) The sonication of the unsheared particles produces a similar particle size reduction to that of shearing with a tissue homogenizer. It is difficult to quantitatively compare the shear produced by a mixer pump installed in a double-shell tank with that produced by the tissue homogenizer in the laboratory. On a qualitative basis, the mixing pump would be expected to have less mechanical and more hydraulic shearing effect than the tissue homogenizer. Since the particle size distribution studies indicate that (for the AY-101 solids) the breaking up of particle aggregates is the main means of particle size reduction, then the hydraulic shearing

76 FR 78180 - Proposed Modification of Class E Airspace; Douglas, AZ

Science.gov (United States)

2011-12-16

...-1313; Airspace Docket No. 11-AWP-17] Proposed Modification of Class E Airspace; Douglas, AZ AGENCY... action proposes to modify Class E airspace at Bisbee Douglas International Airport, Douglas, AZ... feet above the surface at Douglas, AZ. Additional controlled airspace is necessary to accommodate...

Histone H2A.Z is essential for estrogen receptor signaling

Science.gov (United States)

Gévry, Nicolas; Hardy, Sara; Jacques, Pierre-Étienne; Laflamme, Liette; Svotelis, Amy; Robert, François; Gaudreau, Luc

2009-01-01

Incorporation of H2A.Z into the chromatin of inactive promoters has been shown to poise genes for their expression. Here we provide strong evidence that H2A.Z is incorporated into the promoter regions of estrogen receptor (ERα) target genes only upon gene induction, and that, in a cyclic pattern. Moreover, members of the human H2A.Z-depositing complex, p400, also follow the same gene recruitment kinetics as H2A.Z. Importantly, cellular depletion of H2A.Z or p400 leads to a severe defect in estrogen signaling, including loss of estrogen-specific cell proliferation. We find that incorporation of H2A.Z within TFF1 promoter chromatin allows nucleosomes to adopt preferential positions along the DNA translational axis. Finally, we provide evidence that H2A.Z is essential to allow estrogen-responsive enhancer function. Taken together, our results provide strong mechanistic insight into how H2A.Z regulates ERα-mediated gene expression and provide a novel link between H2A.Z–p400 and ERα-dependent gene regulation and enhancer function. PMID:19515975

Influence of the Composition of the Hank’s Balanced Salt Solution on the Corrosion Behavior of AZ31 and AZ61 Magnesium Alloys

Directory of Open Access Journals (Sweden)

Jakub Tkacz

2017-11-01

Full Text Available The electrochemical corrosion characteristics of AZ31 and AZ61 magnesium alloys were analyzed in terms of potentiodynamic tests and electrochemical impedance spectroscopy. The influence of the solution composition and material surface finish was examined also through the analysis of corrosion products created on the samples’ surface after electrochemical measurements in terms of scanning electron microscopy using energy-dispersive spectroscopy. Obtained data revealed the differences in the response of the magnesium alloys to enriched Hank’s Balanced Salt Solution—HBSS+ (with Mg2+ and Ca2+ ions and Hank’s Balanced Salt Solution—HBSS (without Mg2+ and Ca2+ ions. Both examined alloys exhibited better corrosion resistance from the thermodynamic and kinetic point of view in the enriched HBSS+. AZ61 magnesium alloy reached higher values of polarization resistance than AZ31 magnesium alloy in both the used corrosion solutions. Phosphate-based corrosion products were characteristic for the AZ31 and AZ61 alloys tested in the HBSS (without Mg2+ and Ca2+ ions. The combination of phosphate-based corrosion products and clusters of MgO and Mg(OH2 was typical for the surface of samples tested in the enriched HBSS+ (with Mg2+ and Ca2+ ions. Pitting corrosion attack was observed only in the case of enriched HBSS+.

Electrochemical corrosion behavior of AZ91D alloy in ethylene glycol

International Nuclear Information System (INIS)

Fekry, A.M.; Fatayerji, M.Z.

2009-01-01

The effect of concentration on the corrosion behavior of Mg-based alloy AZ91D was investigated in ethylene glycol-water solutions using electrochemical techniques i.e. potentiodynamic polarization, electrochemical impedance measurements (EIS) and surface examination via scanning electron microscope (SEM) technique. This can provide a basis for developing new coolants for magnesium alloy engine blocks. Corrosion behavior of AZ91D alloy by coolant is important in the automotive industry. It was found that the corrosion rate of AZ91D alloy decreased with increasing concentration of ethylene glycol. For AZ91D alloy in chloride >0.05 M or fluoride 0.05 M for fluoride containing ethylene glycol solution, some inhibition effect has been observed. The corrosion of AZ91D alloy in the blank can be effectively inhibited by addition of 0.05 mM paracetamol that reacts with AZ91D alloy and forms a protective film on the surface at this concentration as confirmed by surface examination.

Corrosion product layers on magnesium alloys AZ31 and AZ61: Surface chemistry and protective ability

Energy Technology Data Exchange (ETDEWEB)

Feliu, S., E-mail: sfeliu@cenim.csic.es; Llorente, I.

2015-08-30

Highlights: • Surface chemistry of the corrosion product layers on magnesium alloys. • Influence of the type of alloy on the carbonate surface enrichment. • Relation between surface composition and protection properties. - Abstract: This paper studies the chemical composition of the corrosion product layers formed on magnesium alloys AZ31 and AZ61 following immersion in 0.6 M NaCl, with a view to better understanding their protective action. Relative differences in the chemical nature of the layers were quantified by X-ray photoelectron spectroscopy (XPS), scanning electron microscopy (SEM), energy dispersive analysis of X-ray (EDX) and low-angle X-ray diffraction (XRD). Corrosion behavior was investigated by Electrochemical Impedance Spectroscopy (EIS) and hydrogen evolution measurement. An inhibitive effect from the corrosion product layers was observed from EIS, principally in the case of AZ31, as confirmed by hydrogen evolution tests. A link was found between carbonate enrichment observed by XPS in the surface of the corrosion product layer, concomitant with the increase in the protective properties observed by EIS.

Corrosion product layers on magnesium alloys AZ31 and AZ61: Surface chemistry and protective ability

International Nuclear Information System (INIS)

Feliu, S.; Llorente, I.

2015-01-01

Highlights: • Surface chemistry of the corrosion product layers on magnesium alloys. • Influence of the type of alloy on the carbonate surface enrichment. • Relation between surface composition and protection properties. - Abstract: This paper studies the chemical composition of the corrosion product layers formed on magnesium alloys AZ31 and AZ61 following immersion in 0.6 M NaCl, with a view to better understanding their protective action. Relative differences in the chemical nature of the layers were quantified by X-ray photoelectron spectroscopy (XPS), scanning electron microscopy (SEM), energy dispersive analysis of X-ray (EDX) and low-angle X-ray diffraction (XRD). Corrosion behavior was investigated by Electrochemical Impedance Spectroscopy (EIS) and hydrogen evolution measurement. An inhibitive effect from the corrosion product layers was observed from EIS, principally in the case of AZ31, as confirmed by hydrogen evolution tests. A link was found between carbonate enrichment observed by XPS in the surface of the corrosion product layer, concomitant with the increase in the protective properties observed by EIS

Analysis of Induced Gas Releases During Retrieval of Hanford Double-Shell Tank Waste

International Nuclear Information System (INIS)

Wells, Beric E.; Cuta, Judith M.; Hartley, Stacey A.; Mahoney, Lenna A.; Meyer, Perry A.; Stewart, Charles W.

2002-01-01

Radioactive waste is scheduled to be retrieved from Hanford double-shell tanks AN-103, AN-104, AN-105, and AW-101 to the vitrification plant beginning about 2009. Retrieval may involve decanting the supernatant liquid and/or mixing the waste with jet pumps. In these four tanks, which contain relatively large volumes of retained gas, both of these operations are expected to induce buoyant displacement gas releases that can potentially raise the tank headspace hydrogen concentration to very near the lower flammability limit. This report describes the theory and detailed physical models for both the supernatant decant and jet mixing processes and presents the results from applying the models to these operations in the four tanks. The technical bases for input parameter distributions are elucidated

241-SY-101 multi-functional instrument tree acceptance for beneficial use (ABU)

International Nuclear Information System (INIS)

Erhart, M.F.

1995-01-01

This document formally demonstrates that the ABU process for the 241-SY-101 risers 17B and 17C Multi-functional Instrument Trees (MIT's) has been properly completed in accordance with the approved ABU checklists. For each item required on the ABU Checklist, a bibliography of the documentation prepared and released to satisfy the requirements is provided. Release of this documentation signifies that the tank farm Operations, Engineering, and Maintenance organizations have accepted responsibility for the MIT'S in 241-SY-101 Risers 17B and 17C

Reorganization of Damaged Chromatin by the Exchange of Histone Variant H2A.Z-2

Energy Technology Data Exchange (ETDEWEB)

Nishibuchi, Ikuno [Department of Cellular Biology, Research Institute for Radiation Biology and Medicine, Hiroshima University, Hiroshima (Japan); Department of Radiation Oncology, Graduate School of Biomedical and Health Sciences, Hiroshima University, Hiroshima (Japan); Department of Radiation Oncology, Hiroshima Prefectural Hospital, Hiroshima (Japan); Suzuki, Hidekazu; Kinomura, Aiko; Sun, Jiying; Liu, Ning-Ang [Department of Cellular Biology, Research Institute for Radiation Biology and Medicine, Hiroshima University, Hiroshima (Japan); Horikoshi, Yasunori [Department of Cellular Biology, Research Institute for Radiation Biology and Medicine, Hiroshima University, Hiroshima (Japan); Research Center for Mathematics of Chromatin Live Dynamics, Hiroshima University, Hiroshima (Japan); Shima, Hiroki [Department of Biochemistry, Graduate School of Medical Sciences, Tohoku University, Sendai (Japan); Kusakabe, Masayuki; Harata, Masahiko [Laboratory of Molecular Biology, Graduate School of Agricultural Science, Tohoku University, Sendai (Japan); Fukagawa, Tatsuo [Department of Molecular Genetics, National Institute of Genetics and The Graduate University for Advanced Studies, Mishima (Japan); Ikura, Tsuyoshi [Laboratory of Chromatin Regulatory Network, Department of Mutagenesis, Radiation Biology Center, Kyoto University, Kyoto (Japan); Ishida, Takafumi [Department of Cardiovascular Medicine, Graduate School of Biomedical and Health Sciences, Hiroshima University, Hiroshima (Japan); Nagata, Yasushi [Department of Radiation Oncology, Graduate School of Biomedical and Health Sciences, Hiroshima University, Hiroshima (Japan); Tashiro, Satoshi, E-mail: ktashiro@hiroshima-u.ac.jp [Department of Cellular Biology, Research Institute for Radiation Biology and Medicine, Hiroshima University, Hiroshima (Japan); Research Center for Mathematics of Chromatin Live Dynamics, Hiroshima University, Hiroshima (Japan)

2014-07-15

Purpose: The reorganization of damaged chromatin plays an important role in the regulation of the DNA damage response. A recent study revealed the presence of 2 vertebrate H2A.Z isoforms, H2A.Z-1 and H2A.Z-2. However, the roles of the vertebrate H2A.Z isoforms are still unclear. Thus, in this study we examined the roles of the vertebrate H2A.Z isoforms in chromatin reorganization after the induction of DNA double-strand breaks (DSBs). Methods and Materials: To examine the dynamics of H2A.Z isoforms at damaged sites, we constructed GM0637 cells stably expressing each of the green fluorescent protein (GFP)-labeled H2A.Z isoforms, and performed fluorescence recovery after photobleaching (FRAP) analysis and inverted FRAP analysis in combination with microirradiation. Immunofluorescence staining using an anti-RAD51 antibody was performed to study the kinetics of RAD51 foci formation after 2-Gy irradiation of wild-type (WT), H2A.Z-1- and H2A.Z-2-deficient DT40 cells. Colony-forming assays were also performed to compare the survival rates of WT, H2A.Z-1-, and H2A.Z-2-deficient DT40 cells with control, and H2A.Z-1- and H2A.Z-2-depleted U2OS cells after irradiation. Results: FRAP analysis revealed that H2A.Z-2 was incorporated into damaged chromatin just after the induction of DSBs, whereas H2A.Z-1 remained essentially unchanged. Inverted FRAP analysis showed that H2A.Z-2 was released from damaged chromatin. These findings indicated that H2A.Z-2 was exchanged at DSB sites immediately after the induction of DSBs. RAD51 focus formation after ionizing irradiation was disturbed in H2A.Z-2-deficient DT40 cells but not in H2A.Z-1-deficient cells. The survival rate of H2A.Z-2-deficient cells after irradiation was lower than those of WT and H2A.Z-1- DT40 cells. Similar to DT40 cells, H2A.Z-2-depleted U2OS cells were also radiation-sensitive compared to control and H2A.Z-1-depleted cells. Conclusions: We found that vertebrate H2A.Z-2 is involved in the regulation of the DNA

Reorganization of Damaged Chromatin by the Exchange of Histone Variant H2A.Z-2

International Nuclear Information System (INIS)

Nishibuchi, Ikuno; Suzuki, Hidekazu; Kinomura, Aiko; Sun, Jiying; Liu, Ning-Ang; Horikoshi, Yasunori; Shima, Hiroki; Kusakabe, Masayuki; Harata, Masahiko; Fukagawa, Tatsuo; Ikura, Tsuyoshi; Ishida, Takafumi; Nagata, Yasushi; Tashiro, Satoshi

2014-01-01

Purpose: The reorganization of damaged chromatin plays an important role in the regulation of the DNA damage response. A recent study revealed the presence of 2 vertebrate H2A.Z isoforms, H2A.Z-1 and H2A.Z-2. However, the roles of the vertebrate H2A.Z isoforms are still unclear. Thus, in this study we examined the roles of the vertebrate H2A.Z isoforms in chromatin reorganization after the induction of DNA double-strand breaks (DSBs). Methods and Materials: To examine the dynamics of H2A.Z isoforms at damaged sites, we constructed GM0637 cells stably expressing each of the green fluorescent protein (GFP)-labeled H2A.Z isoforms, and performed fluorescence recovery after photobleaching (FRAP) analysis and inverted FRAP analysis in combination with microirradiation. Immunofluorescence staining using an anti-RAD51 antibody was performed to study the kinetics of RAD51 foci formation after 2-Gy irradiation of wild-type (WT), H2A.Z-1- and H2A.Z-2-deficient DT40 cells. Colony-forming assays were also performed to compare the survival rates of WT, H2A.Z-1-, and H2A.Z-2-deficient DT40 cells with control, and H2A.Z-1- and H2A.Z-2-depleted U2OS cells after irradiation. Results: FRAP analysis revealed that H2A.Z-2 was incorporated into damaged chromatin just after the induction of DSBs, whereas H2A.Z-1 remained essentially unchanged. Inverted FRAP analysis showed that H2A.Z-2 was released from damaged chromatin. These findings indicated that H2A.Z-2 was exchanged at DSB sites immediately after the induction of DSBs. RAD51 focus formation after ionizing irradiation was disturbed in H2A.Z-2-deficient DT40 cells but not in H2A.Z-1-deficient cells. The survival rate of H2A.Z-2-deficient cells after irradiation was lower than those of WT and H2A.Z-1- DT40 cells. Similar to DT40 cells, H2A.Z-2-depleted U2OS cells were also radiation-sensitive compared to control and H2A.Z-1-depleted cells. Conclusions: We found that vertebrate H2A.Z-2 is involved in the regulation of the DNA

Project W-320 Tank 106-C waste retrieval study analysis session report

International Nuclear Information System (INIS)

Bailey, J.W.

1998-01-01

This supporting document has been prepared to make the Kaiser Engineers Hanford Company Project W-320 Tank 106-C Waste Retrieval Study Analysis Session Report readily retrievable. This facilitated session was requested by Westinghouse Hanford Company (WHC) to review the characterization data and select the best alternatives for a double-shell receiver tank and for a sluicing medium for Tank 106-C waste retrieval. The team was composed of WHC and Kaiser Engineers Hanford Company (KEH) personnel knowledgeable about tank farm operations, tank 106-C requirements, tank waste characterization and analysis, and chemical processing. This team was assembled to perform a structured decision analysis evaluation and recommend the best alternative-destination double-shell tank between tanks 101-AY and 102-AY, and the best alternative sluicing medium among dilute complexant (DC), dilute noncomplexant (DNC), and water. The session was facilitated by Richard Harrington and Steve Bork of KEH and was conducted at the Bookwalter Winery in Richland from 7:30 a.m. to 4:00 p.m. from July 27 through July 29, 1993. Attachment 1 (Scope Statement Sheet) identifies the team members, scope, objectives, and deliverables for the session

Evaluation of mitigation strategies in Facility Group 1 double-shell flammable-gas tanks at the Hanford Site

International Nuclear Information System (INIS)

Unal, C.; Sadasivan, P.; Kubic, W.L.; White, J.R.

1997-11-01

Radioactive nuclear waste at the Hanford Site is stored in underground waste storage tanks at the site. The tanks fall into two main categories: single-shell tanks (SSTs) and double-shell tanks (DSTs). There are a total of 149 SSTs and 28 DSTs. The wastes stored in the tanks are chemically complex. They basically involve various sodium salts (mainly nitrite, nitrate, carbonates, aluminates, and hydroxides), organic compounds, heavy metals, and various radionuclides, including cesium, strontium, plutonium, and uranium. The waste is known to generate flammable gas (FG) [hydrogen, ammonia, nitrous oxide, hydrocarbons] by complex chemical reactions. The process of gas generation, retention, and release is transient. Some tanks reach a quasi-steady stage where gas generation is balanced by the release rate. Other tanks show continuous cycles of retention followed by episodic release. There currently are 25 tanks on the Flammable Gas Watch List (FGWL). The objective of this report is to evaluate possible mitigation strategies to eliminate the FG hazard. The evaluation is an engineering study of mitigation concepts for FG generation, retention, and release behavior in Tanks SY-101, AN-103, AN 104, An-105, and Aw-101. Where possible, limited quantification of the effects of mitigation strategies on the FG hazard also is considered. The results obtained from quantification efforts discussed in this report should be considered as best-estimate values. Results and conclusions of this work are intended to help in establishing methodologies in the contractor's controls selection analysis to develop necessary safety controls for closing the FG unreviewed safety question. The general performance requirements of any mitigation scheme are discussed first

49 CFR 172.331 - Bulk packagings other than portable tanks, cargo tanks, tank cars and multi-unit tank car tanks.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Bulk packagings other than portable tanks, cargo tanks, tank cars and multi-unit tank car tanks. 172.331 Section 172.331 Transportation Other Regulations... packagings other than portable tanks, cargo tanks, tank cars and multi-unit tank car tanks. (a) Each person...

Enhancing Microstructure and Mechanical Properties of AZ31-MWCNT Nanocomposites through Mechanical Alloying

Directory of Open Access Journals (Sweden)

J. Jayakumar

2013-01-01

Full Text Available Multiwall carbon nanotubes (MWCNTs reinforced Mg alloy AZ31 nanocomposites were fabricated by mechanical alloying and powder metallurgy technique. The reinforcement material MWCNTs were blended in three weight fractions (0.33%, 0.66%, and 1% with the matrix material AZ31 (Al-3%, zinc-1% rest Mg and blended through mechanical alloying using a high energy planetary ball mill. Specimens of monolithic AZ31 and AZ31-MWCNT composites were fabricated through powder metallurgy technique. The microstructure, density, hardness, porosity, ductility, and tensile properties of monolithic AZ31 and AZ31-MWCNT nano composites were characterized and compared. The characterization reveals significant reduction in CNT (carbon nanoTube agglomeration and enhancement in microstructure and mechanical properties due to mechanical alloying through ball milling.

Electrochemical corrosion behavior of AZ91D alloy in ethylene glycol

Energy Technology Data Exchange (ETDEWEB)

Fekry, A.M. [Chemistry Department, Faculty of Science, Cairo University, Giza 12613 (Egypt)], E-mail: hham4@hotmail.com; Fatayerji, M.Z. [Chemistry Department, Faculty of Science, Cairo University, Giza 12613 (Egypt)

2009-11-01

The effect of concentration on the corrosion behavior of Mg-based alloy AZ91D was investigated in ethylene glycol-water solutions using electrochemical techniques i.e. potentiodynamic polarization, electrochemical impedance measurements (EIS) and surface examination via scanning electron microscope (SEM) technique. This can provide a basis for developing new coolants for magnesium alloy engine blocks. Corrosion behavior of AZ91D alloy by coolant is important in the automotive industry. It was found that the corrosion rate of AZ91D alloy decreased with increasing concentration of ethylene glycol. For AZ91D alloy in chloride >0.05 M or fluoride <0.05 M containing 30% ethylene glycol solution, they are more corrosive than the blank (30% ethylene glycol-70% water). However, at concentrations <0.05 for chloride or >0.05 M for fluoride containing ethylene glycol solution, some inhibition effect has been observed. The corrosion of AZ91D alloy in the blank can be effectively inhibited by addition of 0.05 mM paracetamol that reacts with AZ91D alloy and forms a protective film on the surface at this concentration as confirmed by surface examination.

Comparison of organic and inorganic ion exchangers for removal of cesium and strontium from simulated and actual Hanford 241-AW-101 DSSF tank waste

International Nuclear Information System (INIS)

Brown, G.N.; Bray, L.A.; Carlson, C.D.

1996-04-01

A number of organic and inorganic exchangers are being developed and evaluated for cesium removal from Hanford tank wastes. The exchangers of interest that are investigated in this work include powdered (IONSIV reg-sign IE-910; referred to as IE-910) and engineered (IONSIV reg-sign IE-911; referred to as IE-911) forms of the crystalline silico-titanate (CST) inorganic sorbent developed by Sandia National Laboratories (SNL)/Texas A and M and prepared by UOP; a phenol-formaldehyde (CS-100) resin developed by Rohm and Haas; a resorcinol-formaldehyde (R-F) polymer developed at the Westinghouse Savannah River Company (WSRC) and produced by Boulder Scientific; an inorganic zeolite exchanger produced by UOP (IONSIV reg-sign TIE-96; referred to as TIE-96); an inorganic sodium titanate produced by Allied Signal/Texas A and M (NaTi); and a macrocyclic organic resin developed and produced by IBC Advanced Technologies (SuperLig reg-sign 644; referred to as SL-644). Several of these materials are still under development and may not be in the optimal form. The work described in this report involves the direct comparison of the ion exchange materials for the pretreatment of actual and simulated Hanford tank waste. Data on the performance of all of the exchangers with simulated and actual double shell slurry feed (DSSF) is included. The DSSF waste is a mixture of the supernate from tanks 101-AW (70%), 106-AP (20%) and 102-AP (10%). The comparative parameters include radionuclide removal efficiency under a variety of conditions and material properties (e.g., bed density and percent removable water). Cesium and strontium distribution (K d ), lambda (λ = K d x ρ b ), and decontamination factors (DF) are compared as a function of exchanger contact duration, solution composition (Na and Cs concentration), exchanger/waste phase ratio, and multiple sequential contacts

ANALYSIS RESULTS FOR BUILDING 241 702-AZ A TRAIN

International Nuclear Information System (INIS)

DUNCAN JB; FRYE JM; COOKE CA; LI SW; BROCKMAN FJ

2006-01-01

This report presents the analyses results for three samples obtained under RPP-PLAN-28509, Sampling and Analysis Plan for Building 241 702-AZ A Train. The sampling and analysis was done in response to problem evaluation request number PER-2004-6139, 702-AZ Filter Rooms Need Radiological Cleanup Efforts

ANALYSIS RESULTS FOR BUILDING 241 702-AZ A TRAIN

Energy Technology Data Exchange (ETDEWEB)

DUNCAN JB; FRYE JM; COOKE CA; LI SW; BROCKMAN FJ

2006-12-13

This report presents the analyses results for three samples obtained under RPP-PLAN-28509, Sampling and Analysis Plan for Building 241 702-AZ A Train. The sampling and analysis was done in response to problem evaluation request number PER-2004-6139, 702-AZ Filter Rooms Need Radiological Cleanup Efforts.

Influence of artificial biological fluid composition on the biocorrosion of potential orthopedic Mg-Ca, AZ31, AZ91 alloys

International Nuclear Information System (INIS)

Gu, X N; Zheng, Y F; Chen, L J

2009-01-01

The electrochemical behavior of potential orthopedic Mg-Ca, AZ31 and AZ91 alloys was studied in Hank's solution, Dulbecco's Modified Eagle's Medium (DMEM) and serum-containing medium (DMEM adding 10% fetal bovine serum (DMEM+FBS)) over a 7 day immersion period. The biocorrosion of the above three alloys for various immersion time intervals was investigated by linear polarization and electrochemical impedance spectroscopy (EIS). After 7 day immersion, potentiodynamic polarization tests were carried out and the surface morphologies of experimental samples were examined by scanning electron microscopy (SEM) observation complemented by energy-disperse spectrometer (EDS) analysis. It was shown that the corrosion of magnesium alloys was influenced by the composition of the solution. The results indicated that chloride ion could reduce the corrosion resistance and the hydrocarbonate ions could induce rapid surface passivation. The adsorbed amino acid on the experimental magnesium alloys' surface increased their polarization resistance and reduced current densities. The influence of the serum protein on corrosion was found to be associated with the magnesium alloy compositions. A Mg-Ca alloy exhibited an increased corrosion rate in the presence of serum protein. An AZ31 alloy showed an increased corrosion rate in DMEM+FBS in the initial 3 day immersion and the corrosion rate decreased thereafter. An AZ91 alloy, with high Al content, showed a reduced corrosion rate with the addition of FBS into DMEM.

Influence of artificial biological fluid composition on the biocorrosion of potential orthopedic Mg-Ca, AZ31, AZ91 alloys.

Science.gov (United States)

Gu, X N; Zheng, Y F; Chen, L J

2009-12-01

The electrochemical behavior of potential orthopedic Mg-Ca, AZ31 and AZ91 alloys was studied in Hank's solution, Dulbecco's Modified Eagle's Medium (DMEM) and serum-containing medium (DMEM adding 10% fetal bovine serum (DMEM+FBS)) over a 7 day immersion period. The biocorrosion of the above three alloys for various immersion time intervals was investigated by linear polarization and electrochemical impedance spectroscopy (EIS). After 7 day immersion, potentiodynamic polarization tests were carried out and the surface morphologies of experimental samples were examined by scanning electron microscopy (SEM) observation complemented by energy-disperse spectrometer (EDS) analysis. It was shown that the corrosion of magnesium alloys was influenced by the composition of the solution. The results indicated that chloride ion could reduce the corrosion resistance and the hydrocarbonate ions could induce rapid surface passivation. The adsorbed amino acid on the experimental magnesium alloys' surface increased their polarization resistance and reduced current densities. The influence of the serum protein on corrosion was found to be associated with the magnesium alloy compositions. A Mg-Ca alloy exhibited an increased corrosion rate in the presence of serum protein. An AZ31 alloy showed an increased corrosion rate in DMEM+FBS in the initial 3 day immersion and the corrosion rate decreased thereafter. An AZ91 alloy, with high Al content, showed a reduced corrosion rate with the addition of FBS into DMEM.

Waste tank safety program annual status report for FY 1993, Task 5: Toxicology and epidemiology

International Nuclear Information System (INIS)

Mahlum, D.D.; Young, J.Y.

1993-09-01

A toxicology team independently reviewed analytical data and provided advice concerning potential health effects associated with exposure to tank-vapor constituents at the Hanford site. Most of the emphasis was directed toward Tank 241-C-103, but a preliminary assessment was also made of the toxicologic implication of the cyanide levels in the headspace of Tank 241-C-108. The objectives of this program are to (1) review procedures used for sampling vapors from various tanks, (2) identify constituents in tank-vapor samples that could be related to symptoms reported by waste-tank workers, (3) evaluate the toxicologic implications of those constituents by comparison to established toxicologic data bases, (4) provide advice for additional analytical efforts, and (5) support other activities as requested by the project manager and the cognizant Westinghouse Hanford Company Tank Vapor Issues Safety Resolution Manager

101-SY Dome pressure issues surrounding mitigation pump decontamination during removal

International Nuclear Information System (INIS)

Shaw, S.W.

1995-01-01

This document addresses issues related to use of the spraywands and ring used to decontaminate the mitigation pump installed in 101-SY. It has been determined that use of the wands will influence tank dome pressures as a function of ventilation system configuration, spray drop size, rinse water temperature, and rate at which spraywand flows are established

Grain refinement of AZ31 magnesium alloy by electromagnetic ...

Indian Academy of Sciences (India)

Low-frequency electromagnetic field; AZ31 magnesium alloy; Al4C3; grain refinement. Abstract. The effects of electromagnetic stirring and Al4C3 grain refiner on the grain refinement of semicontinuously cast AZ31 magnesium alloy were discussed in this investigation. The results indicate that electromagnetic stirring has an ...

Bioetanol de la caña de azúcar

Directory of Open Access Journals (Sweden)

N. Aguilar-Rivera

2007-01-01

Full Text Available La búsqueda de la viabilidad y desarrollo de la agroindustria de la caña de azúcar es ya una política imperativa. La producción de caña y azúcar constituyen la principal agroindustria del país, con presencia en 57 ingenios azucareros, localizados en 15 estados de la república. Sin embargo, la actual ineficiencia productiva, altos costos y otras causas —como la entrada de jarabe de maíz de alta fructosa (HFCS de menor costo— han desplazado el consumo interno de azúcar de caña, principalmente en bebidas embotelladas. Todos estos factores han contribuido a la actual y prolongada crisis del sector azucarero. Por lo tanto, la industria debe reestructurarse y reducir la producción de azúcar hacia la competitividad, con la alternativa del etanol combustible en la búsqueda de un futuro para la industria azucarera en México. Con la producción de etanol (EtOH a gran escala, por parte de los ingenios azucareros, es posible visualizar un nuevo y futuro mercado de uso de energías renovables en México, oportunidades para la exportación e incrementar los beneficios ambientales y económicos a través de su uso. La flexibilidad de la producción combinada, de azúcar y etanol en los ingenios, permitirá alcanzar esquemas y objetivos ambientales como nación. Para la industria azucarera, su mayor competidesafío, en este sentido, representa disminuir los costos actuales de producción de etanol a través del uso de la ingeniería genética, uso de diferentes materias primas y la totalidad de la planta de caña de azúcar e introducir mejoras tecnológicas.

Tank vapor characterization project. Headspace vapor characterization of Hanford waste Tank SX-101: Results from samples collected on 07/21/95

International Nuclear Information System (INIS)

Evans, J.C.; Clauss, T.W.; McVeety, B.D.; Pool, K.H.; Olsen, K.B.; Fruchter, J.S.; Silvers, K.L.

1996-05-01

Results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank-farm operations. They include air concentrations of inorganic and organic analytes and grouped compounds from samples. The vapor concentrations are based either on whole-volume samples or on sorbent traps exposed to sample flow. No immediate notifications were needed because analytical results indicated no specific analytes exceeded notification levels. Summary of results: NH3, 3.8 ppmv; NO2, 0.10 ppmv; NO, 0.13 ppm; H2O, 11.8 mg/L; CO2, 338 ppmv; CO, 3 ; methanol, 0.060 ppmv; acetone, 0.033 ppmv; trichlorofluoromethane, 0.023 ppmv; and acetone, 0.034 ppmv

Texture characterisation of hexagonal metals: Magnesium AZ91 alloy, welded by laser processing

International Nuclear Information System (INIS)

Kouadri, A.; Barrallier, L.

2006-01-01

Cooled and cast magnesium AZ91 alloy was welded using a CO 2 laser. The changes in the microstructure were analysed by optical and scanning electron microscopy and X-ray diffraction. Modification of the anisotropic properties was evaluated by the characterization of the texture in the base metal, in the core of the welded zone and in the welded zone close to the surface. In the two former zones, we have not observed a texture. Laser welding only leads to a change of the grain size and a disappearance of the eutectic phase. By contrast, in the welded zone close to the surface, the laser process leads both to a finer microstructure, to a loss of the Al-content and to the presence of several texture components. In this zone, our results showed that these textures are on pyramidal {101-bar 1} and prismatic {101-bar 0} planes. Much of the explanation for such texture rests with the fact that during the laser welding, material solidifies in strong non-equilibrium conditions. The kinetics of the nucleation and the growth are partly controlled by the high-rise and high fall of the temperature and the power produced by the laser process. The nature of the texture has been explained by the presence of a columnar to equiaxed transition in the welded zone

Influence of artificial biological fluid composition on the biocorrosion of potential orthopedic Mg-Ca, AZ31, AZ91 alloys

Energy Technology Data Exchange (ETDEWEB)

Gu, X N; Zheng, Y F [State Key Laboratory for Turbulence and Complex System and Department of Advanced Materials and Nanotechnology, College of Engineering, Peking University, Beijing 100871 (China); Chen, L J, E-mail: yfzheng@pku.edu.c [School of Material Science and Engineering, Shengyang University of Technology, Shenyang 110023 (China)

2009-12-15

The electrochemical behavior of potential orthopedic Mg-Ca, AZ31 and AZ91 alloys was studied in Hank's solution, Dulbecco's Modified Eagle's Medium (DMEM) and serum-containing medium (DMEM adding 10% fetal bovine serum (DMEM+FBS)) over a 7 day immersion period. The biocorrosion of the above three alloys for various immersion time intervals was investigated by linear polarization and electrochemical impedance spectroscopy (EIS). After 7 day immersion, potentiodynamic polarization tests were carried out and the surface morphologies of experimental samples were examined by scanning electron microscopy (SEM) observation complemented by energy-disperse spectrometer (EDS) analysis. It was shown that the corrosion of magnesium alloys was influenced by the composition of the solution. The results indicated that chloride ion could reduce the corrosion resistance and the hydrocarbonate ions could induce rapid surface passivation. The adsorbed amino acid on the experimental magnesium alloys' surface increased their polarization resistance and reduced current densities. The influence of the serum protein on corrosion was found to be associated with the magnesium alloy compositions. A Mg-Ca alloy exhibited an increased corrosion rate in the presence of serum protein. An AZ31 alloy showed an increased corrosion rate in DMEM+FBS in the initial 3 day immersion and the corrosion rate decreased thereafter. An AZ91 alloy, with high Al content, showed a reduced corrosion rate with the addition of FBS into DMEM.

241-SY-101 strain concentration factor development via nonlinear analysis. Volume 1 of 1

International Nuclear Information System (INIS)

1997-01-01

The 241-SY-101 waste storage tank at the Hanford-Site has been known to accumulate and release significant quantities of hydrogen gas. An analysis was performed to assess the tank's structural integrity when subjected to postulated hydrogen deflagration loads. The analysis addressed many nonlinearities and appealed to a strain-based failure criteria. The model used to predict the global response of the tank was not refined enough to confidently predict local peak strains. Strain concentration factors were applied at structural discontinuities that were based on steel-lined reinforced-concrete containment studies. The discontinuities included large penetrations, small penetrations, springline geometries, stud/liner connections, and the 1/2 inch to 3/8 inch liner thickness transition. The only tank specific strain concentration factor applied in the evaluation was for the 1/2 inch to 3/8 inch liner thickness change in the dome. Review of the tank drawings reveals the possibility that a 4 inches Sch. 40 pipe penetrates the dome thickness transition region. It is not obvious how to combine the strain concentration factors for a small penetration with that of a thickness transition to arrive at a composite strain concentration factor. It is the goal of this effort to make an approximate determination of the relative significance of the 4 inch penetration and the 1/2 inch to 3/8 inch thickness transition in the 241-SY-101 dome geometry. This is accomplished by performing a parametric study with three general finite-element models. The first represents the thickness transition only, the second represents a 4 inch penetration only, and the third combines the thickness transition with a penetration model

A summary description of the flammable gas tank safety program

International Nuclear Information System (INIS)

Johnson, G.D.; Sherwood, D.J.

1994-10-01

Radioactive liquid waste may produce hydrogen as result of the interaction of gamma radiation and water. If the waste contains organic chelating agents, additional hydrogen as well as nitrous oxide and ammonia may be produced by thermal and radiolytic decomposition of these organics. Several high-level radioactive liquid waste storage tanks, located underground at the Hanford Site in Washington State, are on a Flammable Gas Watch List. Some contain waste that produces and retains gases until large quantities of gas are released rapidly to the tank vapor space. Tanks nearly-filled to capacity have relatively little vapor space; therefore if the waste suddenly releases a large amount of hydrogen and nitrous oxide, a flammable gas mixture could result. The most notable example of a Hanford waste tank with a flammable gas problem is tank 241-SY-101. Upon occasion waste stored in this tank has released enough flammable gas to burn if an ignition source had been present inside of the tank. Several, other Hanford waste tanks exhibit similar behavior although to a lesser magnitude. Because this behavior was hot adequately-addressed in safety analysis reports for the Hanford Tank Farms, an unreviewed safety question was declared, and in 1990 the Flammable Gas Tank Safety Program was established to address this problem. The purposes of the program are a follows: (1) Provide safety documents to fill gaps in the safety analysis reports, and (2) Resolve the safety issue by acquiring knowledge about gas retention and release from radioactive liquid waste and developing mitigation technology. This document provides the general logic and work activities required to resolve the unreviewed safety question and the safety issue of flammable gas mixtures in radioactive liquid waste storage tanks

Hanford waste tanks-light at the end of the tunnel

International Nuclear Information System (INIS)

POPPITI, J.A.

1999-01-01

The U.S. Department of Energy (DOE) faced several problems in its Hanford Site tank farms in the early nineties. It had 177 waste tanks, ranging in size from 55,000 to 1,100,000 gallons, which contained more than 55 million gallons of liquid and solid high-level radioactive waste (HLW) from a variety of processes. Unfortunately, waste transfer records were incomplete. Chemical reactions going on in the tanks were not totally understood. Every tank had high concentrations of powerful oxidizers in the form of nitrates and nitrites, and some tanks had relatively high concentrations of potential fuels that could react explosively with oxidizers. A few of these tanks periodically released large quantities of hydrogen and nitrous oxide, a mixture that was potentially more explosive than hydrogen and air. Both the nitrate/fuel and hydrogen/nitrous oxide reactions had the potential to rupture a tank exposing workers and the general public to unacceptably large quantities of radioactive material. One tank (241-C-106) was generating so much heat that water had to be added regularly to avoid thermal damage to the tank's concrete exterior shell. The tanks contained more than 250 million Curies of radioactivity. Some of that radioactivity was in the form of fissile plutonium, which represented a potential criticality problem. As awareness of the potential hazards grew, the public and various regulatory agencies brought increasing pressure on DOE to quantify the hazards and mitigate any that were found to be outside accepted risk guidelines. In 1990, then Representative, now Senator Ron Wyden (D-Oregon), introduced an amendment to Public Law 101-510, Section 3137, that required DOE to identify Hanford tanks that might have a serious potential for release of high-level waste

Influence of the Composition of the Hank’s Balanced Salt Solution on the Corrosion Behavior of AZ31 and AZ61 Magnesium Alloys

Czech Academy of Sciences Publication Activity Database

Tkacz, J.; Slouková, K.; Minda, J.; Drábiková, J.; Fintová, Stanislava; Doležal, P.; Wasserbauer, J.

2017-01-01

Roč. 7, č. 11 (2017), č. článku 465. ISSN 2075-4701 Institutional support: RVO:68081723 Keywords : magnesium alloy * AZ31 * AZ61 * HBSS * HBSS+ * EIS * potentiodynamic test Subject RIV: JK - Corrosion ; Surface Treatment of Materials OBOR OECD: Coating and films Impact factor: 1.984, year: 2016 http://www.mdpi.com/2075-4701/7/11/465

Microstructural Characteristics and Mechanical Properties of 2205/AZ31B Laminates Fabricated by Explosive Welding

Directory of Open Access Journals (Sweden)

Yan Li

2017-04-01

Full Text Available A bimetal composite of 2205 duplex stainless steel and AZ31B magnesium alloy was cladded successfully through the method of explosive welding. The microstructural characteristics and mechanical properties of 2205/AZ31B bimetal composite are discussed. The interface of 2205/AZ31B bimetallic composite was a less regular wavy morphology with locally melted pockets. Adiabatic shear bands occurred only in the AZ31B side near explosive welding interface. The microstructure observed with EBSD showed a strong refinement near the interface zones. Line scan confirmed that the interface had a short element diffusion zone which would contribute to the metallurgical bonding between 2205 duplex stainless steel and AZ31B magnesium alloy. The value of micro-hardness near the bonding interface of composite plate increased because of work hardening and grain refinement. The tensile shear strength of bonding interface of 2205/AZ31B composite was 105.63 MPa. Tensile strength of 2205/AZ31B composite material was higher than the base AZ31B. There were two abrupt drops in stress in the stress–strain curves of the 2205/AZ31B composite materials.

Corrosion Evaluation of INTEC Waste Storage Tank WM-182

International Nuclear Information System (INIS)

Dirk, W. J.; Anderson, P. A.

1999-01-01

Irradiated nuclear fuel has been stored and reprocessed at the Idaho National Engineering and Environmental Laboratory since 1953 using facilities located at the Idaho Nuclear Technology and Engineering Center (INTEC). This reprocessing produced radioactive liquid waste which was stored in the Tank Farm. The INTEC Tank Farm consists of eleven vaulted 300,000-gallon underground tanks including Tank WM-182. Tank WM-182 was put into service in 1955, has been filled four times, and has contained aluminum and zirconium fuel reprocessing wastes as well as sodium bearing waste. A program to monitor corrosion in the waste tanks was initiated in 1953 when the first of the eleven Tank Farm tanks was placed in service. Austenitic stainless steel coupons representative of the materials of construction of the tanks are used to monitor internal tank corrosion. This report documents the final inspection of the WM-182 corrosion coupons. Physical examination of the welded corrosion test coupons exposed to the tank bottom conditions of Tank WM-182 revealed very light uniform corrosion. Examination of the external surfaces of the extruded pipe samples showed very light uniform corrosion with slight indications of preferential attack parallel to extrusion marks and start of end grain attack of the cut edges. These indications were only evident when examined under stereo microscope at magnifications of 20X and above. There were no definite indications of localized corrosion, such as cracking, pitting, preferential weld attack, or weld heat affected zone attack on either the welded or extruded coupons. Visual examination of the coupon support cables, where they were not encased in plastic, failed to reveal any indication of liquid-liquid interface attack of any crevice corrosion. Based on the WM-182 coupon evaluations, which have occurred throughout the life of the tank, the metal loss from the tank wall due to uniform corrosion is not expected to exceed 5.5 x 10-1 mil (0.00 055 inch

Preliminary fire hazards analysis for W-211, Initial Tank Retrieval Systems

International Nuclear Information System (INIS)

Huckfeldt, R.A.

1995-01-01

A fire hazards analysis (FHA) was performed for Project W-211, Initial Tank Retrieval System (ITRS), at the Department of Energy (DOE) Hanford site. The objectives of this FHA was to determine (1) the fire hazards that expose the Initial Tank Retrieval System or are inherent in the process, (2) the adequacy of the fire-safety features planned, and (3) the degree of compliance of the project with specific fire safety provisions in DOE orders and related engineering codes and standards. The scope included the construction, the process hazards, building fire protection, and site wide fire protection. The results are presented in terms of the fire hazards present, the potential extent of fire damage, and the impact on employees and public safety. This study evaluated the ITRS with respect to its use at Tank 241-SY-101 only

Test plan for evaluation of primary exhaust ventilation flow meters for double shell hydrogen watch list tanks

International Nuclear Information System (INIS)

Willingham, W.E.

1996-01-01

This document is a plan for testing four different flow meters for use in the primary exhaust ventilation ducts of Double Shell Tanks on the hydrogen watch list that do not already have this capability. This currently includes tanks 241-AW-101, 241-AN-103, 241-AN-104, 241-AN-105, and 241-SY-103. The anticipated airflow velocity in these tanks range from 0.25 m/s(50 ft/min) to 1.78 m/s (350 ft/min). Past experiences at Hanford are forcing the evaluation and selection of instruments to be used at the low flow and relatively high humidity conditions found in these tanks. Based on the results of this test, a flow meter shall be chosen for installation in the primary exhaust ventilation ducts of the above mentioned waste tanks

AZ91C magnesium alloy modified by Cd

DEFF Research Database (Denmark)

Shabadi, R.; Ambat, Rajan; Dwarakadasa, E.S.

2014-01-01

In the present work, the effect of Cd on the microstructure, mechanical properties and general corrosion behaviour of AZ91C alloys was investigated. Addition of Cd was found not to be efficient in modifying/refining the microstructure or β-phase. A morphology change in β-phase from fine continuous...... precipitates to discontinuous β-phase upon the addition of Cd was observed. A marginal increment in mechanical properties was observed. General corrosion behaviour was followed with weight loss measurements, potentiostatic polarisation studies and surface studies in 3.5% sodium chloride solution and 3.......5% sodium chloride with 2% potassium dichromate solution. Cd addition deteriorated the corrosion behaviour of AZ91C. This behaviour was attributed to the formation of chunks of β-phase upon the addition of Cd. AZ91C with refined β-phase distribution, performed rather better in the NaCl solutions....

76 FR 18378 - Amendment of Class E Airspace; Taylor, AZ

Science.gov (United States)

2011-04-04

...-1189; Airspace Docket No. 10-AWP-19] Amendment of Class E Airspace; Taylor, AZ AGENCY: Federal Aviation Administration (FAA), DOT. ACTION: Final rule. SUMMARY: This action will amend Class E airspace at Taylor Airport, Taylor, AZ, to accommodate aircraft using the CAMBO One Departure, and the Area Navigation (RNAV...

System Design Description for the SY-101 Hydrogen Mitigation Test Project Data Acquisition and Control System (DACS-1)

Energy Technology Data Exchange (ETDEWEB)

ERMI, A.M.

2000-01-24

This document describes the hardware and software of the computer subsystems for the Data Acquisition and Control System (DACS) used in mitigation tests conducted on waste tank 241-SY-101 at the Hanford Nuclear Reservation.

Effect of mechanical vibrations on the wear behavior of AZ91 Mg alloy

Science.gov (United States)

Chaturvedi, V.; Pandel, U.; Sharma, A.

2018-02-01

AZ91 Mg alloy is the most promising alloy used for structural applications. The vibration induced methods are effective and economic viable in term of mechanical properties. Sliding wear tests were performed on AZ91 Mg alloy using a pin-on- disc configuration. Wear rates were measured at 5 N and 10N at a sliding velocity of 1m/s for varied frequency within the range of 5- 25Hz and a constant amplitude of 2mm. Microstructures of worn surfaces and wear debris were characterized by field emission scanning electron microscopy (FESEM). It is observed that wear resistance of vibrated AZ91 alloy at 15Hz frequency ad 2mm amplitude was superior than cast AZ91 Mg alloy. Finer grain size and equiaxed grain shape both are important parameters for better wear resistance in vibrated AZ91 Mg alloys. FESEM analysis revealed that wear is considerably affected due to frictional heat generated by the relative motion between AZ91 Mg alloy and EN31 steel surface. No single mechanism was responsible for material loss.

49 CFR 172.330 - Tank cars and multi-unit tank car tanks.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Tank cars and multi-unit tank car tanks. 172.330..., TRAINING REQUIREMENTS, AND SECURITY PLANS Marking § 172.330 Tank cars and multi-unit tank car tanks. (a... material— (1) In a tank car unless the following conditions are met: (i) The tank car must be marked on...

Corrosion behaviour and in vitro/in vivo biocompatibility of surface-modified AZ31 alloy; Comportamiento frente a la corrosion y biocompatibilidad in vitrolin vivo de la aleacion AZ31 modificada superficialmente

Energy Technology Data Exchange (ETDEWEB)

Carboneras, M.; Iglesias, C.; Perez-Maceda, B. T.; Valle, J. A. de; Garcia-Alonso, M. C.; Alobera, M. A.; Clemente, C.; Rubio, J. C.; Escudero, M. I.; Lozano, R. M.

2011-07-01

The present work evaluates the corrosion behaviour and the in vitro/in vivo biocompatibility of the AZ31 magnesium alloy, which fulfills the mechanical requirements of bone. The corrosion kinetic of as-received AZ31 alloy was not compatible with the cell growth. To improve its performance, the AZ31 alloy was surface modified by a chemical conversion treatment in hydrofluoric acid. The magnesium fluoride layer generated by the surface treatment of AZ31 alloy enhances its corrosion behaviour, allowing the in vitro growth of osteoblastic cells over the surface and the in vivo formation of a highly compact layer of new bone tissue. These results lead to consider the magnesium fluoride coating as necessary for potential use of the AZ31 alloy as biodegradable and absorbable implant for bone repair. (Author) 18 refs.

Influence of cerium on passivity behavior of wrought AZ91 alloy

International Nuclear Information System (INIS)

Wang Henan; Li Ying; Wang Fuhui

2008-01-01

In this paper, more focus had been put on the passivity behavior of wrought AZ91 alloy with 1.5 mass% Ce. The passive current density of wrought AZ91 alloy increased with addition of Ce and the passive film became unstable. Structures and compositions of the passive films formed on wrought AZ91 alloy without and with Ce in 0.01 M NaOH aqueous solution were analyzed by potentiostatic polarization, potentiostatic-galvanostatic (P-G) transient technique and X-ray photoelectron spectroscopy (XPS). The results showed that Ce only accumulated in the inner layer of passive film in the form of CeO 2 . Further analysis revealed that there were two main effects of Ce on the passive process of wrought AZ91 alloy: first, the existence of CeO 2 in inner layer of passive film made mass transport through the passive film follow tangent hyperbolic (T) impedance instead of Warburg impedance (W); second, donor concentration (N d ) of the passive film increased by a factor 10 when 1.5 mass% Ce existed in wrought AZ91 alloy, which led to the higher passivity current density

Effect of aluminum coatings on corrosion properties of AZ31 magnesium alloy

Energy Technology Data Exchange (ETDEWEB)

Chiu Liuho; Lin Hsingan; Chen Chunchin; Yang Chihfu [Dept. of materials engineering, Tatung Univ., Taipei (Taiwan); Chang Chiahua; Wu Jenchin [Physical chemistry section, chemical systems research div., Chung-Shan Inst. of Science and Technology, Tao-Yuan (Taiwan)

2003-07-01

This investigation aimed to increase the corrosion resistance of an AZ31 magnesium alloy by an aluminum arc spray coating and a post-treatment consisted of hot pressing and anodizing. It was found that the aluminum arc spraying alone was incapable of protection against corrosion due to the high amount of pores present in the coating layer. In order to solve the problem, densification of the Al arc-sprayed layer was carried out by hot pressing the coated AZ31 Mg alloy plate under an appropriate range of temperature, time and pressure. After hot pressing the Al coated AZ31 Mg alloy plate exhibited a much improved corrosion resistance. A final anodizing treatment applied to the AZ31 alloy with the dense Al coating further improved its resisting to corrosion. The results showed that, by adopting the Al arc spraying, hot pressing and anodizing process, the corrosion current density of the AZ31 alloy in a 3.5 wt% NaCl solution was from 2.1 x 10{sup -6} A/cm{sup 2} (original AZ31) to 3.7 x 10{sup -7} A/cm{sup 2} (after the surface treatment), which value is close to that of an anodized aluminum plate. (orig.)

Microstructure and Aging Behavior of Nonflammable AZ91D Mg Alloy

OpenAIRE

Seok Hong Min; Tae Kwon Ha

2014-01-01

Phase equilibria of AZ91D Mg alloys for nonflammable use, containing Ca and Y, were carried out by using FactSage® and FTLite database, which revealed that solid solution treatment could be performed at temperatures from 400 to 450oC. Solid solution treatment of AZ91D Mg alloy without Ca and Y was successfully conducted at 420oC and supersaturated microstructure with all beta phase resolved into matrix was obtained. In the case of AZ91D Mg alloy with some Ca and Y; howeve...

Tank 50H Tetraphenylborate Destruction Results

International Nuclear Information System (INIS)

Peters, T.B.

2003-01-01

obstacles upon returning Tank 50H to HLW service. The concerns include the potential for retention of flammable gases, nuclear criticality safety implications, and possible combustible solids formation. A recent document describes the initial results of that work

Corrosion properties and corrosion evolution of as-cast AZ91 alloy with rare earth yttrium

International Nuclear Information System (INIS)

Luo, T.J.; Yang, Y.S.

2011-01-01

Highlights: → Minor addition of Y will increase the corrosion resistance of AZ91 alloy, and 0.3 wt.% Y is the optimum addition. → A film composed of Mg(OH) 2 , MgCO 3 , Al(OH) 3 and Al 2 O 3 is formed on the surface of AZ91 alloy with rare earth Y free. → The film of AZ91 alloy with 0.3 wt.% Y is mainly composed of Mg(OH) 2 and MgCO 3 without any Al(OH) 3 and Al 2 O 3 . → The relative quantity of MgCO 3 in the surface film of AZ91 + 0.3 wt.% Y is bigger than that of AZ91 alloy with Y free. → Y 2 O 3 phase is found in the surface film of alloy III, which benefits to stabilize the surface film. -- Abstract: The corrosion resistance property and the corrosion evolution of as-cast AZ91 alloy with rare earth Y addition are investigated by using immersion tests, electrochemical impedance spectroscopy (EIS), and X-ray photoelectron spectroscopy (XPS). The results show that the proper amount of Y in the alloys can improve the corrosion resistance of AZ91 alloys effectively. With the increment of Y, the corrosion rate of the modified AZ91 alloys by Y addition was markedly less than that of AZ91 alloy. The corrosion rate of AZ91 alloy with 0.3 wt.% Y was the slightest, but further addition of Y content over 0.3 wt.% make the corrosion heavier. The XPS analysis suggests that the compound film of AZ91 alloy with 0.3 wt.% Y is mainly composed of Mg(OH) 2 and MgCO 3 without any Al(OH) 3 and Al 2 O 3 , in addition, Y 2 O 3 phase is found in the compound film of AZ91 alloy with 0.3 wt.% Y, which benefits to stabilize the surface film.

Effect of the chemistry and structure of the native oxide surface film on the corrosion properties of commercial AZ31 and AZ61 alloys

Energy Technology Data Exchange (ETDEWEB)

Feliu, Sebastian, E-mail: sfeliu@cenim.csic.es [Centro Nacional de Investigaciones Metalurgicas CSIC, Avda. Gregorio del Amo 8, 28040 Madrid (Spain); Maffiotte, C. [CIEMAT-DT edificio 30, Avda. Complutense, 22, 28040 Madrid (Spain); Samaniego, A.; Galvan, Juan Carlos [Centro Nacional de Investigaciones Metalurgicas CSIC, Avda. Gregorio del Amo 8, 28040 Madrid (Spain); Barranco, Violeta [Centro Nacional de Investigaciones Metalurgicas CSIC, Avda. Gregorio del Amo 8, 28040 Madrid (Spain); Instituto de Ciencias de Materiales de Madrid, ICMM, Consejo Superior de Investigaciones Cientificas, CSIC, Sor Juana Ines de la Cruz, 3, Cantoblanco, 28049, Madrid (Spain)

2011-08-01

The purpose of this study has been to advance in knowledge of the chemical composition, structure and thickness of the thin native oxide film formed spontaneously in contact with the laboratory atmosphere on the surface of freshly polished commercial AZ31 and AZ61 alloys with a view to furthering the understanding of protection mechanisms. For comparative purposes, and to more fully describe the behaviour of the native oxide film, the external oxide films formed as a result of the manufacturing process (as-received condition) have been characterised. The technique applied in this research to study the thin oxide films (thickness of just a few nanometres) present on the surface of the alloys has basically been XPS (X-ray photoelectron spectroscopy) in combination with ion sputtering. Corrosion properties of the alloys were studied in 0.6 M NaCl by measuring charge transfer resistance values, which are deduced from EIS (electrochemical impedance spectroscopy) measurements after 1 h of exposure. Alloy AZ61 generally showed better corrosion resistance than AZ31, and the freshly polished alloys showed better corrosion resistance than the alloys in as-received condition. This is attributed to a combination of (1) higher thickness of the native oxide film on the AZ61 alloy and (2) greater uniformity of the oxide film in the polished condition. The formation of an additional oxide layer composed by a mixture of spinel (MgAl{sub 2}O{sub 4}) and MgO seems to diminish the protective properties of the passive layer on the surface of the alloys in as-received condition.

Effect of the chemistry and structure of the native oxide surface film on the corrosion properties of commercial AZ31 and AZ61 alloys

International Nuclear Information System (INIS)

Feliu, Sebastian; Maffiotte, C.; Samaniego, A.; Galvan, Juan Carlos; Barranco, Violeta

2011-01-01

The purpose of this study has been to advance in knowledge of the chemical composition, structure and thickness of the thin native oxide film formed spontaneously in contact with the laboratory atmosphere on the surface of freshly polished commercial AZ31 and AZ61 alloys with a view to furthering the understanding of protection mechanisms. For comparative purposes, and to more fully describe the behaviour of the native oxide film, the external oxide films formed as a result of the manufacturing process (as-received condition) have been characterised. The technique applied in this research to study the thin oxide films (thickness of just a few nanometres) present on the surface of the alloys has basically been XPS (X-ray photoelectron spectroscopy) in combination with ion sputtering. Corrosion properties of the alloys were studied in 0.6 M NaCl by measuring charge transfer resistance values, which are deduced from EIS (electrochemical impedance spectroscopy) measurements after 1 h of exposure. Alloy AZ61 generally showed better corrosion resistance than AZ31, and the freshly polished alloys showed better corrosion resistance than the alloys in as-received condition. This is attributed to a combination of (1) higher thickness of the native oxide film on the AZ61 alloy and (2) greater uniformity of the oxide film in the polished condition. The formation of an additional oxide layer composed by a mixture of spinel (MgAl 2 O 4 ) and MgO seems to diminish the protective properties of the passive layer on the surface of the alloys in as-received condition.

76 FR 30299 - Proposed Establishment of Class E Airspace; Kayenta, AZ

Science.gov (United States)

2011-05-25

...-0393; Airspace Docket No. 11-AWP-2] Proposed Establishment of Class E Airspace; Kayenta, AZ AGENCY... action proposes to establish Class E Airspace at Kayenta Airport, Kayenta, AZ. Controlled airspace is...) standard instrument approach procedures at Kayenta Airport. The FAA is proposing this action to enhance the...

EnviroAtlas - Phoenix, AZ - Block Groups

Data.gov (United States)

U.S. Environmental Protection Agency — This EnviroAtlas dataset is the base layer for the Phoenix, AZ EnviroAtlas area. The block groups are from the US Census Bureau and are included/excluded based on...

Thermal and combined thermal and radiolytic reactions involving nitrous oxide, hydrogen, nitrogen, and ammonia in contact with tank 241-SY-101 simulated waste

International Nuclear Information System (INIS)

Bryan, S.A.; Pederson, L.R.

1996-02-01

Work described in this report was conducted at Pacific Northwest National Laboratory (PNNL) for the Flammable Gas Safety Project, the purpose of which is to develop information needed to support Westinghouse Hanford Company (WHC) in their efforts to ensure the safe interim storage of wastes at the Hanford Site. Described in this report are the results of tests to evaluate the rates of thermal and combined thermal and radiolytic reactions involving flammable gases in the presence of Tank 241-SY-101 simulated waste. Flammable gases generated by the radiolysis of water and by the thermal and radiolytic decomposition of organic waste constituents may themselves participate in further reactions. Examples include the decomposition of nitrous oxide to yield nitrogen and oxygen, the reaction of nitrous oxide and hydrogen to produce nitrogen and water, and the reaction of nitrogen and hydrogen to produce ammonia. The composition of the gases trapped in bubbles in the wastes might therefore change continuously as a function of the time that the gas bubbles are retained

Project W-151 Tank 101-AZ Waste Retrieval System Year 2000 Compliance Assessment Project Plan

International Nuclear Information System (INIS)

BUSSELL, J.H.

1999-01-01

This document contains a limited assessment of Year 2000 compliance for Project W-151. Additional information is provided as a road map to project documents and other references that may be used to verify Year 2000 compliance

Differential deposition of H2A.Z in rice seedling tissue during the day-night cycle.

Science.gov (United States)

Zhang, Kang; Xu, Wenying; Wang, Chunchao; Yi, Xin; Su, Zhen

2017-03-04

Chromatin structure has an important role in modulating gene expression. The incorporation of histone variants into the nucleosome leads to important changes in the chromatin structure. The histone variant H2A.Z is highly conserved between different species of fungi, animals, and plants. However, dynamic changes to H2A.Z in rice have not been reported during the day-night cycle. In this study, we generated genome wide maps of H2A.Z for day and night time in harvested seedling tissues by combining chromatin immunoprecipitation and high-throughput sequencing. The analysis results for the H2A.Z data sets detected 7099 genes with higher depositions of H2A.Z in seedling tissues harvested at night compared with seedling tissues harvested during the day, whereas 4597 genes had higher H2A.Z depositions in seedlings harvested during the day. The gene expression profiles data suggested that H2A.Z probably negatively regulated gene expression during the day-night cycle and was involved in many important biologic processes. In general, our results indicated that H2A.Z may play an important role in plant responses to the diurnal oscillation process.

A study on the relationships between corrosion properties and chemistry of thermally oxidised surface films formed on polished commercial magnesium alloys AZ31 and AZ61

Science.gov (United States)

Feliu, Sebastián; Samaniego, Alejandro; Barranco, Violeta; El-Hadad, A. A.; Llorente, Irene; Serra, Carmen; Galván, J. C.

2014-03-01

This paper studies the changes in chemical composition of the thin oxide surface films induced by heating in air at 200 °C for time intervals from 5 min to 60 min on the freshly polished commercial AZ31 and AZ61 alloys with a view to better understanding their protective properties. This thermal treatment resulted in the formation of layers enriched in metallic aluminium at the interface between the outer MgO surface films and the bulk material. A strong link was found between the degree of metallic Al enrichment in the subsurface layer (from 10 to 15 at.%) observed by XPS (X-ray photoelectron spectroscopy) in the AZ61 treated samples and the increase in protective properties observed by EIS (electrochemical impedance spectroscopy) in the immersion test in 0.6 M NaCl. Heating for 5-60 min in air at 200 °C seems to be an effective, easy to perform and inexpensive method for increasing the corrosion resistance of the AZ61 alloy by approximately two or three times.

The A-Z of electronuclear power

International Nuclear Information System (INIS)

Temmerman, G. de; Chedorge, D.

1992-01-01

Both a bilingual translation aid and a reference book, 'The A-Z of Electronuclear Power' covers the full range of electronuclear activities and the English and French terminology relating to the different phases of the modern industrial process

76 FR 3570 - Proposed Amendment of Class E Airspace; Taylor, AZ

Science.gov (United States)

2011-01-20

...-1189; Airspace Docket No. 10-AWP-19] Proposed Amendment of Class E Airspace; Taylor, AZ AGENCY: Federal... proposes to modify Class E airspace at Taylor Airport, Taylor, AZ. Controlled airspace is necessary to accommodate aircraft using the CAMBO One Departure Area Navigation (RNAV) out of Taylor Airport. The FAA is...

AX Tank Farm tank removal study

Energy Technology Data Exchange (ETDEWEB)

SKELLY, W.A.

1999-02-24

This report examines the feasibility of remediating ancillary equipment associated with the 241-AX Tank Farm at the Hanford Site. Ancillary equipment includes surface structures and equipment, process waste piping, ventilation components, wells, and pits, boxes, sumps, and tanks used to make waste transfers to/from the AX tanks and adjoining tank farms. Two remedial alternatives are considered: (1) excavation and removal of all ancillary equipment items, and (2) in-situ stabilization by grout filling, the 241-AX Tank Farm is being employed as a strawman in engineering studies evaluating clean and landfill closure options for Hanford single-shell tanks. This is one of several reports being prepared for use by the Hanford Tanks Initiative Project to explore potential closure options and to develop retrieval performance evaluation criteria for tank farms.

AX Tank Farm tank removal study

International Nuclear Information System (INIS)

SKELLY, W.A.

1998-01-01

This report considers the feasibility of exposing, demolishing, and removing underground storage tanks from the 241-AX Tank Farm at the Hanford Site. For the study, it was assumed that the tanks would each contain 360 ft 3 of residual waste (corresponding to the one percent residual Inventory target cited in the Tri-Party Agreement) at the time of demolition. The 241-AX Tank Farm is being employed as a ''strawman'' in engineering studies evaluating clean and landfill closure options for Hanford single-shell tank farms. The report is one of several reports being prepared for use by the Hanford Tanks Initiative Project to explore potential closure options and to develop retrieval performance evaluation criteria for tank farms

Electrochemical polymerization of pyrrole over AZ31 Mg alloy for biomedical applications

International Nuclear Information System (INIS)

Srinivasan, A.; Ranjani, P.; Rajendran, N.

2013-01-01

Highlights: ► Polymerization of pyrrole over AZ31 Mg was carried out using cyclic voltammetry. ► Pyrrole concentration was optimized to accomplish the adherent and uniform coating. ► Effect of monomer concentration on the surface morphology was discussed. ► Corrosion resistance of AZ31 Mg in SBF was studied as a function of Py concentration. ► PPy coated AZ31 Mg alloy exhibited enhanced corrosion resistance at 0.25 M of Py. -- Abstract: Electrochemical polymerization of pyrrole (Py) from aqueous salicylate solution over AZ31 Mg alloy was carried out using cyclic voltammetry (CV). The effect of monomer concentration on the surface and electrochemical corrosion in simulated body fluid (SBF) were analysed. Attenuated total reflection-infrared (ATR-IR) spectra showed the characteristic ring stretching peaks for polypyrrole (PPy). Scanning electron microscopy (SEM) and atomic force microscopy (AFM) studies exhibited typical cauliflower morphology with rough surface for PPy coated AZ31 Mg alloy. Open circuit potential measurement and potentiodynamic polarization studies revealed that the coating prepared using 0.25 M of Py had positive shift of about 120 mV in corrosion potential and lower corrosion current density (0.03 mA/cm 2 ) compared to other concentrations and uncoated AZ31 Mg alloy (0.25 mA/cm 2 ). Electrochemical impedance spectroscopic (EIS) studies of uncoated and PPy coated Mg alloy in SBF revealed three-time constants behaviour with about one order of increment in impedance value for 0.25 M of Py

Action plan for response to abnormal conditions in Hanford high level radioactive liquid waste storage tanks containing flammable gases

International Nuclear Information System (INIS)

Sherwood, D.J.

1994-03-01

Radioactive liquid waste tends to produce hydrogen as a result of the interaction of gamma radiation and water. In tanks containing organic chelating agents, additional hydrogen gas as well as nitrous oxide and ammonia can be produced by thermal and radiolytic decomposition of these organics. Several high-level radioactive liquid waste storage tanks, located underground at the Hanford Site, contain waste that retains the gases produced in them until large quantities are released rapidly to the tank vapor space. Tanks filled to near capacity have relatively little vapor space; therefore, if the waste suddenly releases a large amount of hydrogen and nitrous oxide, a flammable gas mixture may result. The most notable waste tank with a flammable gas problem is tank 241-SY-101. Waste in this tank has occasionally released enough flammable gas to burn if an ignition source had been present inside of the tank. Several other waste tanks exhibit similar behavior to a lesser magnitude. Administrative controls have been developed to assure that these Flammable Gas Watch List tanks are safely maintained. Responses have also been developed for off-normal conditions which might develop in these tanks. In addition, scientific and engineering studies are underway to further understand and mitigate the behavior of the Flammable Gas Watch List tanks

Friction welding of AZ31 magnesium alloy

Energy Technology Data Exchange (ETDEWEB)

Tsujino, R.; Ochi, H. [Osaka Inst. of Tech., Osaka (Japan); Kawai, G. [Osaka Sangyo Univ., Osaka (Japan); Yamaguchi, H.; Ogawa, K. [Osaka Prefecture Univ., Osaka (Japan); Suga, Y. [Keio Univ., Kanagawa (Japan)

2003-07-01

In this paper, for an acceleration of utilization of magnesium alloy which is being interested in recent years, friction welding of AZ31 magnesium alloy was carried out, and the joint performance was discussed in relation to the deformation heat input in the upset stage and upset loss as a evaluation factor. Where, the deformation heat input in the upset stage is mechanical work represented by the product of upset speed and axial pressure. As a result, it was made clear that the friction welding of AZ31 magnesium alloy was easy in the atmosphere, and good welded joints without a non- adhesion area at the weld interface could de obtained. Moreover, the evaluation factors discussed were possible to evaluate to joint performance. (orig.)

Tank 241-BY-108 tank characterization plan

International Nuclear Information System (INIS)

Carpenter, B.C.

1994-01-01

The sampling and analytical needs associated with the 51 Hanford Site underground storage tanks classified on one or more of the four Watch Lists (ferrocyanide, organic, flammable gas, and high heat), and the safety screening of all 177 tanks have been identified through the Data Quality Objective (DQO) process. DQOs identity information needed by a program group in the Tank Waste Remediation System concerned with safety issues, regulatory requirements, or the transporting and processing of tank waste. This Tank Characterization Plan will identify characterization objectives for tank BY-108 pertaining to sample collection, sample preparation and analysis, and laboratory analytical evaluation and reporting requirements. In addition, an estimate of the current contents and status of the tank is given. Single-shell tank BY-108 is classified as a Ferrocyanide Watch List tank. The tank was declared an assumed leaker and removed from service in 1972; interim stabilized was completed in February 1985. Although not officially an Organic Watch List tank, restrictions have been placed on intrusive operations by Standing Order number-sign 94-16 (dated 09/08/94) since the tank is suspected to contain or to have contained a floating organic layer

Plasmodium falciparum signal peptide peptidase cleaves malaria heat shock protein 101 (HSP101). Implications for gametocytogenesis

International Nuclear Information System (INIS)

Baldwin, Michael; Russo, Crystal; Li, Xuerong; Chishti, Athar H.

2014-01-01

Highlights: • PfSPP is an ER resident protease. • PfSPP is expressed both as a monomer and dimer. • The signal peptide of HSP101 is the first known substrate of PfSPP. • Reduced PfSPP activity may significantly affect ER homeostasis. - Abstract: Previously we described the identification of a Plasmodium falciparum signal peptide peptidase (PfSPP) functioning at the blood stage of malaria infection. Our studies also demonstrated that mammalian SPP inhibitors prevent malaria parasite growth at the late-ring/early trophozoite stage of intra-erythrocytic development. Consistent with its role in development, we tested the hypothesis that PfSPP functions at the endoplasmic reticulum of P.falciparum where it cleaves membrane-bound signal peptides generated following the enzyme activity of signal peptidase. The localization of PfSPP to the endoplasmic reticulum was confirmed by immunofluorescence microscopy and immunogold electron microscopy. Biochemical analysis indicated the existence of monomer and dimer forms of PfSPP in the parasite lysate. A comprehensive bioinformatics screen identified several candidate PfSPP substrates in the parasite genome. Using an established transfection based in vivo luminescence assay, malaria heat shock protein 101 (HSP101) was identified as a substrate of PfSPP, and partial inhibition of PfSPP correlated with the emergence of gametocytes. This finding unveils the first known substrate of PfSPP, and provides new perspectives for the function of intra-membrane proteolysis at the erythrocyte stage of malaria parasite life cycle

Plasmodium falciparum signal peptide peptidase cleaves malaria heat shock protein 101 (HSP101). Implications for gametocytogenesis

Energy Technology Data Exchange (ETDEWEB)

Baldwin, Michael; Russo, Crystal; Li, Xuerong [Department of Developmental, Molecular and Chemical Biology, Tufts University School of Medicine, Boston, MA 02111 (United States); Chishti, Athar H., E-mail: athar.chishti@tufts.edu [Department of Developmental, Molecular and Chemical Biology, Tufts University School of Medicine, Boston, MA 02111 (United States); Sackler School of Graduate Biomedical Sciences, Programs in Physiology, Pharmacology, and Microbiology, Tufts University School of Medicine, Boston, MA 02111 (United States)

2014-08-08

Highlights: • PfSPP is an ER resident protease. • PfSPP is expressed both as a monomer and dimer. • The signal peptide of HSP101 is the first known substrate of PfSPP. • Reduced PfSPP activity may significantly affect ER homeostasis. - Abstract: Previously we described the identification of a Plasmodium falciparum signal peptide peptidase (PfSPP) functioning at the blood stage of malaria infection. Our studies also demonstrated that mammalian SPP inhibitors prevent malaria parasite growth at the late-ring/early trophozoite stage of intra-erythrocytic development. Consistent with its role in development, we tested the hypothesis that PfSPP functions at the endoplasmic reticulum of P.falciparum where it cleaves membrane-bound signal peptides generated following the enzyme activity of signal peptidase. The localization of PfSPP to the endoplasmic reticulum was confirmed by immunofluorescence microscopy and immunogold electron microscopy. Biochemical analysis indicated the existence of monomer and dimer forms of PfSPP in the parasite lysate. A comprehensive bioinformatics screen identified several candidate PfSPP substrates in the parasite genome. Using an established transfection based in vivo luminescence assay, malaria heat shock protein 101 (HSP101) was identified as a substrate of PfSPP, and partial inhibition of PfSPP correlated with the emergence of gametocytes. This finding unveils the first known substrate of PfSPP, and provides new perspectives for the function of intra-membrane proteolysis at the erythrocyte stage of malaria parasite life cycle.

In Vitro Studies on the Degradability, Bioactivity, and Cell Differentiation of PRP/AZ31B Mg Alloys Composite Scaffold

Directory of Open Access Journals (Sweden)

Jian Zou

2017-01-01

Full Text Available In recent years, more and more methods have been developed to improve the bioactivity of the biodegradable materials in bone tissue regeneration. In present study, we used rat mesenchymal stem cells (rMSCs to evaluate the outcomes of Mg alloys (AZ31B, Magnesium, and Aluminum and Platelet-rich plasma (PRP/Mg alloys on rMSCs biocompatibility and osteogenic differentiation. Water absorption experiments indicated that both bare AZ31B and PRP/AZ31B were capable of absorbing large amounts of water. But the water absorption ratio for PRP/AZ31B was significantly higher than that for bare AZ31B. The degradability experiments implied that both samples degraded at same speed. rMSCs on the surface of AZ31B distributed more and better than those on the AZ31B scaffold. In ALP activity experiment, the activity of rMSCs on the PRP/AZ31B was markedly higher than that on the AZ31B scaffolds on the 7th day and 14th day. qRT-PCR also showed that OPN and OCN were expressed in both samples. OPN and OCN expression in PRP/AZ31B sample were higher than those in bare AZ31B samples. In summary, the in vitro study implied that AZ31B combined with PRP could remarkably improve cell seeding, attachment, proliferation, and differentiation.

Status of test results of electrochemical organic oxidation of a tank 241-SY-101 simulated waste

International Nuclear Information System (INIS)

Colby, S.A.

1994-06-01

This report presents scoping test results of an electrochemical waste pretreatment process to oxidize organic compounds contained in the Hanford Site's radioactive waste storage tanks. Electrochemical oxidation was tested on laboratory scale to destroy organics that are thought to pose safety concerns, using a nonradioactive, simulated tank waste. Minimal development work has been applied to alkaline electrochemical organic destruction. Most electrochemical work has been directed towards acidic electrolysis, as in the metal purification industry, and silver catalyzed oxidation. Alkaline electrochemistry has traditionally been associated with the following: (1) inefficient power use, (2) electrode fouling, and (3) solids handling problems. Tests using a laboratory scale electrochemical cell oxidized surrogate organics by applying a DC electrical current to the simulated tank waste via anode and cathode electrodes. The analytical data suggest that alkaline electrolysis oxidizes the organics into inorganic carbonate and smaller carbon chain refractory organics. Electrolysis treats the waste without adding chemical reagents and at ambient conditions of temperature and pressure. Cell performance was not affected by varying operating conditions and supplemental electrolyte additions

Selectividad de caña de azúcar en bovinos

Directory of Open Access Journals (Sweden)

E. M. Aranda

2009-01-01

Full Text Available Se condujo este trabajo con el objetivo de conocer la aceptación del forraje de caña de azúcar integral en la dieta con base de pasto para bovinos adultos fistulados en rumen y duodeno estabulados; para ello, se ofreció pasto estrella africana (Cynodon plectostachyus y caña de azúcar (Saccharum officinarum mezclada con 1% de urea a voluntad, por separado. Se midió la composición de los alimentos, consumo de materia seca, pH ruminal y duodenal. Se utilizaron técnicas de la estadística descriptiva, como la media aritmética y error estándar. La proteína, paredes celulares y hemicelulosa fueron mayores para el pasto estrella. El consumo de MS fue de 46.6 y 38.1% para la caña de azúcar y el pasto, respectivamente; el pH ruminal tuvo valores de 7.3, disminuyendo a 6.7 a las 12 h.; y el pH de la digesta duodenal osciló de 3.7 a 4.7. Se concluye que hubo mayor aceptación por la caña de azúcar y el pH ruminal no indica valores que afecten la celulólisis ruminal.

41 CFR 101-27.101 - General.

Science.gov (United States)

2010-07-01

... 41 Public Contracts and Property Management 2 2010-07-01 2010-07-01 true General. 101-27.101 Section 101-27.101 Public Contracts and Property Management Federal Property Management Regulations System FEDERAL PROPERTY MANAGEMENT REGULATIONS SUPPLY AND PROCUREMENT 27-INVENTORY MANAGEMENT 27.1-Stock...

Comportamiento mecánico de la aleación AZ31 reforzada con nanofibras de carbono

Directory of Open Access Journals (Sweden)

Adeva, P.

2010-12-01

Full Text Available The aim of this study is to investigate the effect of reinforcing AZ31 with carbon nanofibres. The materials AZ31, AZ31-1 % C y AZ31-2 % C were produced by a conventional powder metallurgy route consisting of mechanical mixing of nanofibres and powders of AZ31, cool compactation and extrusion at 350 °C. After extrusion the three materials exhibited a recrystallized microstructure of similar grain size, fine and rather inhomogeneous. Furthermore, they presented a weak fibre texture with basal plane parallel to the extrusion direction. The tensile properties were affected by the nanofibres presence only at 100 °C. At this temperature, yield strength and tensile strength were 30% higher than in the unreinforced alloy.

En este trabajo se ha estudiado el efecto de la adición de nanofibras de carbono en las propiedades mecánicas de la aleación AZ31 procesada por una ruta pulvimetalúrgica convencional. Se prepararon tres materiales, AZ31, AZ31- 1 % C y AZ31-2 % C. Tras una mezcla mecánica de las nanofibras con los polvos de AZ31, se precompactaron en frío y se extruyeron a 350 °C. Los tres presentan una microestructura recristalizada con un tamaño de grano similar, fino aunque algo heterogéneo. Los tres materiales presentan una débil textura de fibra con el plano basal paralelo a la dirección de extrusión. Las propiedades mecánicas a tracción únicamente se ven afectadas por la presencia de nanofibras a 100 °C superando los materiales reforzados en un 30 % a los valores de límite elástico y resistencia de la aleación sin reforzar.

Microstructure and corrosion resistance of Ce–V conversion coating on AZ31 magnesium alloy

International Nuclear Information System (INIS)

Jiang, Xiao; Guo, Ruiguang; Jiang, Shuqin

2015-01-01

Highlights: • Through simple chemical conversion process, a Ce–V conversion coating is prepared on AZ31 magnesium alloy. The coating (∼2 μm thick) has a duplex structure and is composed of Mg, Al, Ce, V and O in the outer layer and Mg, Al, V, F and O in the inner layer. • The Ce–V conversion coating can increase the E corr by 157 mV and decrease the i corr by 80 times compared to AZ31 magnesium alloy substrate. Moreover, the performance of the Ce–V conversion coating excels the chromate conversion coating on AZ31 magnesium alloy. • The EIS results of Ce–V conversion coating indicate an increase of 10× in the corrosion resistance and a delay in the corrosion process kinetics compared to uncoated AZ31 magnesium alloy in 3.5 wt.% NaCl solution. • The ball cratering is a simple and effective technique of thickness measurement for chemical conversion coating. - Abstract: A Ce–V conversion coating was developed to improve the corrosion resistance of AZ31 magnesium alloy. Scanning electronic microscope (SEM), energy dispersive X-ray spectroscopy (EDS), X-ray photoelectron spectrometer (XPS), grazing incidence X-ray diffraction (GIXRD) and the ball cratering test were adopted to study the morphology, chemical composition, structure and thickness of the coating. The coating has duplex structure with network and its thickness is about 2 μm. The coating contains high contents of Ce and V, which exhibits amorphous structure. Potentiodynamic polarization shows the coating can increase the corrosion potential and reduce the corrosion current density of AZ31 magnesium alloy. Moreover, the electrochemical impedance spectra exhibit the coating significantly improves the corrosion resistance of AZ31 magnesium alloy. Results indicate that the Ce–V conversion coating can provide effective protection to AZ31 magnesium alloy

Microstructure and corrosion resistance of Ce–V conversion coating on AZ31 magnesium alloy

Energy Technology Data Exchange (ETDEWEB)

Jiang, Xiao, E-mail: xiaoxiao217@126.com; Guo, Ruiguang; Jiang, Shuqin

2015-06-30

Highlights: • Through simple chemical conversion process, a Ce–V conversion coating is prepared on AZ31 magnesium alloy. The coating (∼2 μm thick) has a duplex structure and is composed of Mg, Al, Ce, V and O in the outer layer and Mg, Al, V, F and O in the inner layer. • The Ce–V conversion coating can increase the E{sub corr} by 157 mV and decrease the i{sub corr} by 80 times compared to AZ31 magnesium alloy substrate. Moreover, the performance of the Ce–V conversion coating excels the chromate conversion coating on AZ31 magnesium alloy. • The EIS results of Ce–V conversion coating indicate an increase of 10× in the corrosion resistance and a delay in the corrosion process kinetics compared to uncoated AZ31 magnesium alloy in 3.5 wt.% NaCl solution. • The ball cratering is a simple and effective technique of thickness measurement for chemical conversion coating. - Abstract: A Ce–V conversion coating was developed to improve the corrosion resistance of AZ31 magnesium alloy. Scanning electronic microscope (SEM), energy dispersive X-ray spectroscopy (EDS), X-ray photoelectron spectrometer (XPS), grazing incidence X-ray diffraction (GIXRD) and the ball cratering test were adopted to study the morphology, chemical composition, structure and thickness of the coating. The coating has duplex structure with network and its thickness is about 2 μm. The coating contains high contents of Ce and V, which exhibits amorphous structure. Potentiodynamic polarization shows the coating can increase the corrosion potential and reduce the corrosion current density of AZ31 magnesium alloy. Moreover, the electrochemical impedance spectra exhibit the coating significantly improves the corrosion resistance of AZ31 magnesium alloy. Results indicate that the Ce–V conversion coating can provide effective protection to AZ31 magnesium alloy.

Effect of rare earth elements on high cycle fatigue behavior of AZ91 alloy

International Nuclear Information System (INIS)

Mokhtarishirazabad, M.; Boutorabi, S.M.A.; Azadi, M.; Nikravan, M.

2013-01-01

This article investigates effects of adding rare earth elements (RE) into a magnesium–aluminum–zinc alloy (the AZ91 alloy) on its high cycle fatigue (HCF) behavior. For this purpose, AZ91 and AZ91+1% RE (AZE911) alloys were gravity casted in a metallic die. RE elements were added to the AZ91 alloy in the form of mischmetals. Microscopic evaluations with the scanning electron microscopy (SEM) and mechanical tests include tensile, hardness and HCF behaviors, were performed on prepared samples. Rotary bending fatigue tests were carried out at a stress ratio (R) of −1 and a frequency of 125 Hz, at the room temperature, in the air. The microscopic investigation demonstrates that the addition of 1% RE elements leads to the formation of Al 11 RE 3 intermetallic particles which is associated to the reduction of β-(Mg 17 Al 12 ) phases. Results of mechanical experiments suggest a negligible effect of adding 1% RE elements on mechanical properties of the AZ91 alloy. Curves of stress-life (S–N) shows an increase in the fatigue strength at 10 5 cycles, from 100±10 MPa to 135±10 MPa, when RE elements were added to the AZ91 alloy

Effect of rare earth elements on high cycle fatigue behavior of AZ91 alloy

Energy Technology Data Exchange (ETDEWEB)

Mokhtarishirazabad, M., E-mail: mehdi-mokhtari@hotmail.com [School of Metallurgy and Materials Engineering, Iran University of Science and Technology, Tehran (Iran, Islamic Republic of); Boutorabi, S.M.A. [School of Metallurgy and Materials Engineering, Iran University of Science and Technology, Tehran (Iran, Islamic Republic of); Azadi, M.; Nikravan, M. [Irankhodro Powertrain Company (IPCO), Tehran (Iran, Islamic Republic of)

2013-12-10

This article investigates effects of adding rare earth elements (RE) into a magnesium–aluminum–zinc alloy (the AZ91 alloy) on its high cycle fatigue (HCF) behavior. For this purpose, AZ91 and AZ91+1% RE (AZE911) alloys were gravity casted in a metallic die. RE elements were added to the AZ91 alloy in the form of mischmetals. Microscopic evaluations with the scanning electron microscopy (SEM) and mechanical tests include tensile, hardness and HCF behaviors, were performed on prepared samples. Rotary bending fatigue tests were carried out at a stress ratio (R) of −1 and a frequency of 125 Hz, at the room temperature, in the air. The microscopic investigation demonstrates that the addition of 1% RE elements leads to the formation of Al{sub 11}RE{sub 3} intermetallic particles which is associated to the reduction of β-(Mg{sub 17}Al{sub 12}) phases. Results of mechanical experiments suggest a negligible effect of adding 1% RE elements on mechanical properties of the AZ91 alloy. Curves of stress-life (S–N) shows an increase in the fatigue strength at 10{sup 5} cycles, from 100±10 MPa to 135±10 MPa, when RE elements were added to the AZ91 alloy.

Acceptance test report, 241-SY-101 Flexible Receiver System, Phase 3 testing

International Nuclear Information System (INIS)

Ritter, G.A.

1995-01-01

This document summarizes the results of the phase 3 acceptance test of the 241-SY-101 Flexible Receiver System (FRS). The purpose of this acceptance test is to verify the sealing integrity of the FRS to ensure that the release of waste and aerosols will be minimized during the removal of the test mixer pump from Tank 241-SY-101. The FRS is one of six major components of the Equipment Removal System, which has been designed to retrieve, transport, and store the mixer pump. This acceptance test was performed at the 306E Facility in the 300 area from January 10, 1995 to January 17, 1995. The Phase 3 test consisted of two parts. Part one was a water leak test of the seal between the blast shield and mock load distribution frame (LDF) to ensure that significant contamination of the pump pit and waste interaction with the aluminum impact-limiting material under the LDF are prevented during the pump removal operation. The second part of this acceptance test was an air leak test of the assembled flexible receiver system. The purpose of this test was to verify that the release of hazardous aerosols will be minimized if the tank dome pressure becomes slightly positive during the decontamination of the mixer pump

Type B Investigation Report for 241-SY-101 Pump Start and 241-C-106 Pit Cleanout

Energy Technology Data Exchange (ETDEWEB)

Ewalt, J.R.

1993-09-01

In accordance with the direction of the Department of Energy (DOE) Manager, Richland Operations Office, a Type ``B`` investigation in accordance with the DOE Order 5484.1, Environmental Protection, Safety and Health Protection Information Reporting Requirements, has been conducted. The scope of the investigation included two events: The ``Inadvertent Mixer Pump Operation at 241-SY-101`` (RL-WHC-TANK FARM-1993-069); ``Inadequate Work Control Results in Personnel Skin Contamination at 241-C-106, Pit B`` (RL-WHC-TANK FARM-1993-071) events. Additionally, at the request of the President of the WHC, a broader investigation into Waste Tank Farm ``safety practices`` and ``Conduct of Operations`` was also conducted. The review was focused on (1) WHC organizations performing operations, maintenance, and radiological safety tasks; and (2) KEH organizations performing major maintenance tasks.

Test Excavations at Painted Rock Reservoir: Sites AZ Z:1:7, AZ Z:1:8, and AZ S:16:36.

Science.gov (United States)

1981-11-01

were dehydrated, trimmed to 12-mm squares, and 77 78 placed in catalyzed (0.05-percent benzoyl peroxide), deinhibited acrylic monomer (50-percent methyl ...refrigeration for 72 hours in partially polymerized ethyl- methyl methacrylate. The methacrylate was polymerized at 50 degrees C and 50 p.s.i. nitrogen...pollen types from fill of Feature 10, AZ Z:l:8 Pollen Type Percent Pinus spp. 0.5 Cheno-Am 92.0 Gramineae 2.0 Ambrosia spp. 0.5 Malvaceae 0.5 Zea spp. 2.0

Riser configuration, Tank 241-A-105, light duty utility arm

International Nuclear Information System (INIS)

Boucher, T.D.

1994-01-01

The light-duty utility arm (LDUA) is a seven-joint stainless steel robotic arm with a payload capacity of 75 lb. The robotic arm is deployed vertically with a maximum vertical reach of 63 ft. and a maximum horizontal reach of 13.5 ft. The functional requirements of the LDUA system are mapping and characterization of waste in Hanford single-shell tanks (SST) before and during waste retrieval. The LDUA system consists of a mobile deployment system (MDS), a vertical positioning mast (VPM), a tank riser interface confinement (TRIC), the LDUA, and a controller subsystem or support trailer. Currently, the system is in design and is subject to change; however, the LDUA or robotic arm will be deployed through a 12-in. riser above the tank dome. Field trips were performed to gather specifics for future deployment of the LDUA in Tank 241-A-105. The purpose of this report is to support two previous reports for the investigation of SSTs for deployment of the LDUA system. The first report identified the availability of risers while the second report identified the availability of Tanks 241-A-105, 241-A-S-109, 241-A-T-101, and 241-A-T-109 for deployment of the LDUA system. The second report also identified those 4- and 12-in. risers that could be used for deployment of the LDUA and camera system. This report addresses accessibility to the 241-A Tank Farm and the usability of the Tank 241-A-105 risers. The following information for assisting in the design and deployment of the LDUA will be discussed in this report: radiation survey; flange identification; high resolution video; computer simulated model; and field survey

Examples of grid generation with implicitly specified surfaces using GridPro (TM)/az3000. 1: Filleted multi-tube configurations

Science.gov (United States)

Cheng, Zheming; Eiseman, Peter R.

1995-01-01

With examples, we illustrate how implicitly specified surfaces can be used for grid generation with GridPro/az3000. The particular examples address two questions: (1) How do you model intersecting tubes with fillets? and (2) How do you generate grids inside the intersected tubes? The implication is much more general. With the results in a forthcoming paper which develops an easy-to-follow procedure for implicit surface modeling, we provide a powerful means for rapid prototyping in grid generation.

Tank 244A tank characterization plan

International Nuclear Information System (INIS)

Schreiber, R.D.

1994-01-01

The Double-Shell Tank (DST) System currently receives waste from the Single-Shell Tank (SST) System in support of SST stabilization efforts or from other on-site facilities which generate or store waste. Waste is also transferred between individual DSTs. The mixing or commingling of potentially incompatible waste types at the Hanford Site must be addressed prior to any waste transfers into the DSTs. The primary goal of the Waste Compatibility Program is to prevent the formation of an Unreviewed Safety Question (USQ) as a result of improper waste management. Tank 244A is a Double Contained Receiver Tank (DCRT) which serves as any overflow tank for the East Area Farms. Waste material is able to flow freely between the underground storage tanks and tank 244A. Therefore, it is necessary to test the waste in tank 244A for compatibility purposes. Two issues related to the overall problem of waste compatibility must be evaluated: Assurance of continued operability during waste transfer and waste concentration and Assurance that safety problems are not created as a result of commingling wastes under interim storage. The results of the grab sampling activity prescribed by this Tank Characterization Plan shall help determine the potential for four kinds of safety problems: criticality, flammable gas accumulation, energetics, and corrosion and leakage

Kv10.1 potassium channel: from the brain to the tumors.

Science.gov (United States)

Cázares-Ordoñez, V; Pardo, L A

2017-10-01

The KCNH1 gene encodes the Kv10.1 (Eag1) ion channel, a member of the EAG (ether-à-go-go) family of voltage-gated potassium channels. Recent studies have demonstrated that KCHN1 mutations are implicated in Temple-Baraitser and Zimmermann-Laband syndromes and other forms of developmental deficits that all present with mental retardation and epilepsy, suggesting that Kv10.1 might be important for cognitive development in humans. Although the Kv10.1 channel is mainly expressed in the mammalian brain, its ectopic expression occurs in 70% of human cancers. Cancer cells and tumors expressing Kv10.1 acquire selective advantages that favor cancer progression through molecular mechanisms that involve several cellular pathways, indicating that protein-protein interactions may be important for Kv10.1 influence in cell proliferation and tumorigenesis. Several studies on transcriptional and post-transcriptional regulation of Kv10.1 expression have shown interesting mechanistic insights about Kv10.1 role in oncogenesis, increasing the importance of identifying the cellular factors that regulate Kv10.1 expression in tumors.

Analysis of the Deformability of Two-Layer Materials AZ31/Eutectic / Analiza Możliwości Odkształcania Plastycznego Materiału Dwuwarstwowego AZ31/Eutektyka

Directory of Open Access Journals (Sweden)

Mola R.

2015-12-01

Full Text Available The paper present the results of physical simulation of the deformation of the two-layered AZ31/eutectic material using the Gleeble 3800 metallurgical processes simulator. The eutectic layer was produced on the AZ31 substrate using thermochemical treatment. The specimens of AZ31 alloy were heat treated in contact with aluminium powder at 445°C in a vacuum furnace. Depending on the heating time, Al-enriched surface layers with a thickness of 400, 700 and 1100 μm were fabricated on a substrate which was characterized by an eutectic structure composed of the Mg17Al12 phase and a solid solution of aluminium in magnesium. In the study, physical simulation of the fabricated two-layered specimens with a varying thickness of the eutectic layer were deformed using the plane strain compression test at various values of strain rates. The testing results have revealed that it is possible to deform the two-layered AZ31/eutectic material at low strain rates and small deformation values.

Corrosion behaviour of laser surface melted magnesium alloy AZ91D

International Nuclear Information System (INIS)

Taltavull, C.; Torres, B.; Lopez, A.J.; Rodrigo, P.; Otero, E.; Atrens, A.; Rams, J.

2014-01-01

A high power diode laser (HPDL) was used to produce laser surface melting (LSM) treatments on the surface of the Mg alloy AZ91D. Different treatments with different microstructures were produced by varying the laser-beam power and laser-scanning speed. Corrosion evaluation, using hydrogen evolution and electrochemical measurements, led to a relationship between microstructure and corrosion. Most corrosion rates for LSM treated specimens were within the scatter of the as-received AZ91D, whereas some treatments gave higher corrosion rates and some of the samples had corrosion rates lower than the average of the corrosion rate for AZ91D. There were differences in corroded surface morphology. Nevertheless laser treatments introduced surface discontinuities, which masked the effect of the microstructure. Removing these surface defects decreased the corrosion rate for the laser-treated samples. - Highlights: • Corrosion behavior of AZ91D Mg alloys is intimately related with its microstructure. • Laser surface melting treatments allows surface modification of the microstructure. • Different laser parameters can achieve different microstructures. • Controlling laser parameters can produce different corrosion rates and morphologies. • Increase of surface roughness due to laser treatment is relevant to the corrosion rate

Tank 5 Model for Sludge Removal Analysis

International Nuclear Information System (INIS)

LEE, SI

2004-01-01

Computational fluid dynamics methods have been used to develop and provide slurry pump operational guidance for sludge heel removal in Tank 5. Flow patterns calculated by the model were used to evaluate the performance of various combinations of operating pumps and their orientation under steady-state indexed and transient oscillation modes. A model used for previous analyses has been updated to add the valve housing distribution piping and pipe clusters of the cooling coil supply system near pump no. 8 to the previous tank Type-I model. In addition, the updated model included twelve concrete support columns. This model would provide a more accurate assessment of sludge removal capabilities. The model focused on removal of the sludge heel located near the wall of Tank 5 using the two new slurry pumps. The models and calculations were based on prototypic tank geometry and expected normal operating conditions as defined by Tank Closure Project Engineering. Computational fluid dynamics models of Tank 5 with different operating conditions were developed using the FLUENT (trademark) code. The modeling results were used to assess the efficiency of sludge suspension and removal operations in the 75-ft tank. The models employed a three-dimensional approach, a two-equation turbulence model, and an approximate representation of flow obstructions. The calculated local velocity was used as a measure of sludge removal and mixing capability. For the simulations, modeling calculations were performed with indexed pump orientations until an optimum flow pattern near the potential location of the sludge heel was established for sludge removal. The calculated results demonstrated that the existing slurry pumps running at 3801 gpm flowrate per nozzle could remove the sludge from the tank with a 101 in liquid level, based on a historical minimum sludge suspension velocity of 2.27 ft/sec. The only exception is the region within maximum 4.5 ft distance from the tank wall boundary at the

Theoretical comparison between solar combisystems based on bikini tanks and tank-in-tank solar combisystems

DEFF Research Database (Denmark)

Yazdanshenas, Eshagh; Furbo, Simon; Bales, Chris

2008-01-01

Theoretical investigations have shown that solar combisystems based on bikini tanks for low energy houses perform better than solar domestic hot water systems based on mantle tanks. Tank-in-tank solar combisystems are also attractive from a thermal performance point of view. In this paper......, theoretical comparisons between solar combisystems based on bikini tanks and tank-in-tank solar combisystems are presented....

Vibration Considerations for Cryogenic Tanks Using Glass Bubbles Insulation

Science.gov (United States)

Werlink, Rudolph J.; Fesmire, James E.; Sass, Jared P.

2011-01-01

The use of glass bubbles as an efficient and practical thermal insulation system has been previously demonstrated in cryogenic storage tanks. One such example is a spherical, vacuum-jacketed liquid hydrogen vessel of 218,000 liter capacity where the boiloff rate has been reduced by approximately 50 percent. Further applications may include non-stationary tanks such as mobile tankers and tanks with extreme duty cycles or exposed to significant vibration environments. Space rocket launch events and mobile tanker life cycles represent two harsh cases of mechanical vibration exposure. A number of bulk fill insulation materials including glass bubbles, perlite powders, and aerogel granules were tested for vibration effects and mechanical behavior using a custom design holding fixture subjected to random vibration on an Electrodynamic Shaker. The settling effects for mixtures of insulation materials were also investigated. The vibration test results and granular particle analysis are presented with considerations and implications for future cryogenic tank applications. A thermal performance update on field demonstration testing of a 218,000 L liquid hydrogen storage tank, retrofitted with glass bubbles, is presented. KEYWORDS: Glass bubble, perlite, aerogel, insulation, liquid hydrogen, storage tank, mobile tanker, vibration.

Tank 241-BY-111 tank characterization plan

International Nuclear Information System (INIS)

Homi, C.S.

1994-01-01

The sampling and analytical needs associated with the 51 Hanford Site underground storage tanks classified on one or more of the four Watch Lists (ferrocyanide, organic, flammable gas, and high heat), and the safety screening of all 177 tanks have been identified through the Data Quality Objective (DQO) process. DQO's identify information needed by a program group in the Tank Waste Remediation System concerned with safety issues, regulatory requirements, or the transporting and processing of tank waste. This Tank Characterization Plan will identify characterization objectives for Tank BY-111 pertaining to sample collection, sample preparation and analysis, and laboratory analytical evaluation and reporting requirements. In addition, an estimate of the current contents and status of the tank is given

Washing and caustic leaching of Hanford tank sludge: Results of FY 1997 studies

International Nuclear Information System (INIS)

Lumetta, G.J.; Burgeson, I.E.; Wagner, M.J.; Liu, J.; Chen, Y.L.

1997-08-01

The current plan for remediating the Hanford tank farms consists of waste retrieval, pretreatment, treatment (immobilization), and disposal. The tank wastes will be partitioned into high-level and low-level fractions. The HLW will be immobilized in a borosilicate glass matrix; the resulting glass canisters will then be disposed of in a geologic repository. Because of the expected high cost of HLW vitrification and geologic disposal, pretreatment processes will be implemented to reduce the volume of immobilized high-level waste (IHLW). Caustic leaching (sometimes referred to as enhanced sludge washing or ESW) represents the baseline method for pretreating Hanford tank sludges. Caustic leaching is expected to remove a large fraction of the Al, which is present in large quantities in Hanford tank sludges. A significant portion of the P is also expected to be removed from the sludge by metathesis of water-insoluble metal phosphates to insoluble hydroxides and soluble Na 3 PO 4 . Similar metathesis reactions can occur for insoluble sulfate salts, allowing the removal of sulfate from the HLW stream. This report describes the sludge washing and caustic leaching tests performed at the Pacific Northwest National Laboratory in FY 1996. The sludges used in this study were taken from Hanford tanks AN-104, BY-108, S-101, and S-111

A study on the relationships between corrosion properties and chemistry of thermally oxidised surface films formed on polished commercial magnesium alloys AZ31 and AZ61

International Nuclear Information System (INIS)

Feliu, Sebastián; Samaniego, Alejandro; Barranco, Violeta; El-Hadad, A.A.; Llorente, Irene; Serra, Carmen; Galván, J.C.

2014-01-01

Highlights: • Surface chemistry of heat treated magnesium alloys. • Relation between heat treatment and aluminium subsurface enrichment. • Relation between surface composition and corrosion behaviour. - Abstract: This paper studies the changes in chemical composition of the thin oxide surface films induced by heating in air at 200 °C for time intervals from 5 min to 60 min on the freshly polished commercial AZ31 and AZ61 alloys with a view to better understanding their protective properties. This thermal treatment resulted in the formation of layers enriched in metallic aluminium at the interface between the outer MgO surface films and the bulk material. A strong link was found between the degree of metallic Al enrichment in the subsurface layer (from 10 to 15 at.%) observed by XPS (X-ray photoelectron spectroscopy) in the AZ61 treated samples and the increase in protective properties observed by EIS (electrochemical impedance spectroscopy) in the immersion test in 0.6 M NaCl. Heating for 5–60 min in air at 200 °C seems to be an effective, easy to perform and inexpensive method for increasing the corrosion resistance of the AZ61 alloy by approximately two or three times

A study on the relationships between corrosion properties and chemistry of thermally oxidised surface films formed on polished commercial magnesium alloys AZ31 and AZ61

Energy Technology Data Exchange (ETDEWEB)

Feliu, Sebastián, E-mail: sfeliu@cenim.csic.es [Centro Nacional de Investigaciones Metalúrgicas CSIC, Avda. Gregorio del Amo 8, 28040 Madrid (Spain); Samaniego, Alejandro [Centro Nacional de Investigaciones Metalúrgicas CSIC, Avda. Gregorio del Amo 8, 28040 Madrid (Spain); Barranco, Violeta [Instituto de Ciencias de Materiales de Madrid, ICMM, Consejo Superior de Investigaciones Científicas, CSIC, Sor Juana Inés de la Cruz, 3, Cantoblanco, 28049, Madrid (Spain); El-Hadad, A.A. [Physics Department, Faculty of Science, Al-Azhar University, Nasr City 11884, Cairo (Egypt); Llorente, Irene [Centro Nacional de Investigaciones Metalúrgicas CSIC, Avda. Gregorio del Amo 8, 28040 Madrid (Spain); Serra, Carmen [Servicio de Nanotecnologia y Análisis de Superficies, CACTI, Universidade de Vigo, 36310 Vigo (Spain); Galván, J.C. [Centro Nacional de Investigaciones Metalúrgicas CSIC, Avda. Gregorio del Amo 8, 28040 Madrid (Spain)

2014-03-01

Highlights: • Surface chemistry of heat treated magnesium alloys. • Relation between heat treatment and aluminium subsurface enrichment. • Relation between surface composition and corrosion behaviour. - Abstract: This paper studies the changes in chemical composition of the thin oxide surface films induced by heating in air at 200 °C for time intervals from 5 min to 60 min on the freshly polished commercial AZ31 and AZ61 alloys with a view to better understanding their protective properties. This thermal treatment resulted in the formation of layers enriched in metallic aluminium at the interface between the outer MgO surface films and the bulk material. A strong link was found between the degree of metallic Al enrichment in the subsurface layer (from 10 to 15 at.%) observed by XPS (X-ray photoelectron spectroscopy) in the AZ61 treated samples and the increase in protective properties observed by EIS (electrochemical impedance spectroscopy) in the immersion test in 0.6 M NaCl. Heating for 5–60 min in air at 200 °C seems to be an effective, easy to perform and inexpensive method for increasing the corrosion resistance of the AZ61 alloy by approximately two or three times.

Biodegradable behaviors of AZ31 magnesium alloy in simulated body fluid

International Nuclear Information System (INIS)

Song Yingwei; Shan Dayong; Chen Rongshi; Zhang Fan; Han Enhou

2009-01-01

Magnesium alloys have unique advantages to act as biodegradable implants for clinical application. The biodegradable behaviors of AZ31 in simulated body fluid (SBF) for various immersion time intervals were investigated by electrochemical impedance spectroscopy (EIS) tests and scanning electron microscope (SEM) observation, and then the biodegradable mechanisms were discussed. It was found that a protective film layer was formed on the surface of AZ31 in SBF. With increasing of immersion time, the film layer became more compact. If the immersion time was more than 24 h, the film layer began to degenerate and emerge corrosion pits. In the meantime, there was hydroxyapatite particles deposited on the film layer. The hydroxyapatite is the essential component of human bone, which indicates the perfect biocompatibility of AZ31 magnesium alloy.

Underground Storage Tanks - Storage Tank Locations

Data.gov (United States)

NSGIC Education | GIS Inventory — A Storage Tank Location is a DEP primary facility type, and its sole sub-facility is the storage tank itself. Storage tanks are aboveground or underground, and are...

Study on fused/cast AZS refractories for deployment in vitrification of radioactive waste effluents

Energy Technology Data Exchange (ETDEWEB)

Sengupta, Pranesh, E-mail: praneshsengupta@gmail.com [Materials Science Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Mishra, R.K.; Soudamini, N. [Waste Management Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Sen, D.; Mazumder, S. [Solid State Physics Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Kaushik, C.P. [Waste Management Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Ajithkumar, T.G. [Central NMR Facility, National Chemical Laboratory, Pune 411 008 (India); Banerjee, K. [Nuclear Recycle Group, Mumbai 400 085 (India)

2015-12-15

‘Fused/cast Al{sub 2}O{sub 3}–ZrO{sub 2}–SiO{sub 2} (FC-AZS)’ is being considered as ‘glass contact refractory’ within ceramic melters, to be used for nuclear waste immobilization. Microstructural analyses reveal random distributions of baddeleyite (ZrO{sub 2}) within aluminosilicate (Al{sub 2}SiO{sub 5}) matrix. {sup 27}Al and {sup 29}Si NMR data suggest that within aluminosilicate matrix Al occurs in both 4- and 6-fold co-ordinations whereas Si prefers a 4-fold environment. Polydispersity of pores has been studied with small-angle neutron scattering (SANS) technique. Corrosion rates of FC-AZS within 6 M HNO{sub 3}, simulated wastes (500 h exposure), and borosilicate melt (975 °C, 800 h exposure) are found to be 0.38 × 10{sup 3} μmy{sup −1}, 0.13 × 10{sup 3} μmy{sup −1} and 4.75 × 10{sup 3} μmy{sup −1} respectively. A comparison of chemical interaction data clearly suggests that FC-AZS exhibits better chemical durability than AZC refractory (Al{sub 2}O{sub 3}–ZrO{sub 2}–Cr{sub 2}O{sub 3}, also used for similar purpose). Thermal cycling studies indicate that FC-AZS retains structural integrity (including compressive strength and density) even up to 20 cycles. - Highlights: • Vitrification of nuclear waste using AZS refractory within ceramic melter. • Microstructure of AZS refractory. • Interaction of AZS with simulated high level waste and glass.

Vapor characterization of Tank 241-C-103

International Nuclear Information System (INIS)

Huckaby, J.L.; Story, M.S.

1994-06-01

The Westinghouse Hanford Company Tank Vapor Issue Resolution Program has developed, in cooperation with Northwest Instrument Systems, Inc., Oak Ridge National Laboratory, Oregon Graduate Institute of Science and Technology, Pacific Northwest Laboratory, and Sandia National Laboratory, the equipment and expertise to characterize gases and vapors in the high-level radioactive waste storage tanks at the Hanford Site in south central Washington State. This capability has been demonstrated by the characterization of the tank 241-C-103 headspace. This tank headspace is the first, and for many reasons is expected to be the most problematic, that will be characterized (Osborne 1992). Results from the most recent and comprehensive sampling event, sample job 7B, are presented for the purpose of providing scientific bases for resolution of vapor issues associated with tank 241-C-103. This report is based on the work of Clauss et al. 1994, Jenkins et al. 1994, Ligotke et al. 1994, Mahon et al. 1994, and Rasmussen and Einfeld 1994. No attempt has been made in this report to evaluate the implications of the data presented, such as the potential impact of headspace gases and vapors to tank farm workers health. That and other issues will be addressed elsewhere. Key to the resolution of worker health issues is the quantitation of compounds of toxicological concern. The Toxicology Review Panel, a panel of Pacific Northwest Laboratory experts in various areas, of toxicology, has chosen 19 previously identified compounds as being of potential toxicological concern. During sample job 7B, the sampling and analytical methodology was validated for this preliminary list of compounds of toxicological concern. Validation was performed according to guidance provided by the Tank Vapor Conference Committee, a group of analytical chemists from academic institutions and national laboratories assembled and commissioned by the Tank Vapor Issue Resolution Program

Vapor characterization of Tank 241-C-103

Energy Technology Data Exchange (ETDEWEB)

Huckaby, J.L. [Westinghouse Hanford Co., Richland, WA (United States); Story, M.S. [Northwest Instrument Systems, Inc. Richland, WA (United States)

1994-06-01

The Westinghouse Hanford Company Tank Vapor Issue Resolution Program has developed, in cooperation with Northwest Instrument Systems, Inc., Oak Ridge National Laboratory, Oregon Graduate Institute of Science and Technology, Pacific Northwest Laboratory, and Sandia National Laboratory, the equipment and expertise to characterize gases and vapors in the high-level radioactive waste storage tanks at the Hanford Site in south central Washington State. This capability has been demonstrated by the characterization of the tank 241-C-103 headspace. This tank headspace is the first, and for many reasons is expected to be the most problematic, that will be characterized (Osborne 1992). Results from the most recent and comprehensive sampling event, sample job 7B, are presented for the purpose of providing scientific bases for resolution of vapor issues associated with tank 241-C-103. This report is based on the work of Clauss et al. 1994, Jenkins et al. 1994, Ligotke et al. 1994, Mahon et al. 1994, and Rasmussen and Einfeld 1994. No attempt has been made in this report to evaluate the implications of the data presented, such as the potential impact of headspace gases and vapors to tank farm workers health. That and other issues will be addressed elsewhere. Key to the resolution of worker health issues is the quantitation of compounds of toxicological concern. The Toxicology Review Panel, a panel of Pacific Northwest Laboratory experts in various areas, of toxicology, has chosen 19 previously identified compounds as being of potential toxicological concern. During sample job 7B, the sampling and analytical methodology was validated for this preliminary list of compounds of toxicological concern. Validation was performed according to guidance provided by the Tank Vapor Conference Committee, a group of analytical chemists from academic institutions and national laboratories assembled and commissioned by the Tank Vapor Issue Resolution Program.

Tank 241-C-103 tank characterization plan

International Nuclear Information System (INIS)

Schreiber, R.D.

1994-01-01

The data quality objective (DQO) process was chosen as a tool to be used to identify the sampling analytical needs for the resolution of safety issues. A Tank Characterization Plant (TCP) will be developed for each double shell tank (DST) and single-shell tank (SST) using the DQO process. There are four Watch list tank classifications (ferrocyanide, organic salts, hydrogen/flammable gas, and high heat load). These classifications cover the six safety issues related to public and worker health that have been associated with the Hanford Site underground storage tanks. These safety issues are as follows: ferrocyanide, flammable gas, organic, criticality, high heat, and vapor safety issues. Tank C-103 is one of the twenty tanks currently on the Organic Salts Watch List. This TCP will identify characterization objectives pertaining to sample collection, hot cell sample isolation, and laboratory analytical evaluation and reporting requirements in accordance with the appropriate DQO documents. In addition, the current contents and status of the tank are projected from historical information. The relevant safety issues that are of concern for tanks on the Organic Salts Watch List are: the potential for an exothermic reaction occurring from the flammable mixture of organic materials and nitrate/nitrite salts that could result in a release of radioactive material and the possibility that other safety issues may exist for the tank