3718-F Alkali Metal Treatment and Storage Facility Closure Plan

International Nuclear Information System (INIS)

1992-11-01

The Hanford Site, located northwest of the city of Richland, Washington, houses reactors, chemical-separation systems, and related facilities used for the production of special nuclear materials, as well as for activities associated with nuclear energy development. The 300 Area of the Hanford Site contains reactor fuel manufacturing facilities and several research and development laboratories. The 3718-F Alkali Metal Treatment and Storage Facility (3718-F Facility), located in the 300 Area, was used to store and treat alkali metal wastes. Therefore, it is subject to the regulatory requirements for the storage and treatment of dangerous wastes. Closure will be conducted pursuant to the requirements of the Washington Administrative Code (WAC) 173-303-610 (Ecology 1989) and 40 CFR 270.1. Closure also will satisfy the thermal treatment facility closure requirements of 40 CFR 265.381. This closure plan presents a description of the 3718-F Facility, the history of wastes managed, and the approach that will be followed to close the facility. Only hazardous constituents derived from 3718-F Facility operations will be addressed

3718-F Alkali Metal Treatment and Storage Facility Closure Plan

International Nuclear Information System (INIS)

1991-12-01

Since 1987, Westinghouse Hanford Company has been a major contractor to the U.S. Department of Energy-Richland Operations Office and has served as co-operator of the 3718-F Alkali Metal Treatment and Storage Facility, the waste management unit addressed in this closure plan. The closure plan consists of a Part A Dangerous waste Permit Application and a RCRA Closure Plan. An explanation of the Part A Revision (Revision 1) submitted with this document is provided at the beginning of the Part A section. The closure plan consists of 9 chapters and 5 appendices. The chapters cover: introduction; facility description; process information; waste characteristics; groundwater; closure strategy and performance standards; closure activities; postclosure; and references

3718-F Alkali Metal Treatment and Storage Facility Closure Plan

Energy Technology Data Exchange (ETDEWEB)

None

1991-12-01

Since 1987, Westinghouse Hanford Company has been a major contractor to the U.S. Department of Energy-Richland Operations Office and has served as co-operator of the 3718-F Alkali Metal Treatment and Storage Facility, the waste management unit addressed in this closure plan. The closure plan consists of a Part A Dangerous waste Permit Application and a RCRA Closure Plan. An explanation of the Part A Revision (Revision 1) submitted with this document is provided at the beginning of the Part A section. The closure plan consists of 9 chapters and 5 appendices. The chapters cover: introduction; facility description; process information; waste characteristics; groundwater; closure strategy and performance standards; closure activities; postclosure; and references.

2727-S Nonradioactive Dangerous Waste Storage Facility Closure Plan

International Nuclear Information System (INIS)

Wilczek, T.A.; Laws, J.R.; Izatt, R.D.

1992-01-01

This closure plan describes the activities for final closure of the 2727-S Nonradioactive Dangerous Waste Storage (NRDWS) Facility at the Hanford Site. The 2727-S NRDWS Facility provided container storage for nonradioactive dangerous and extremely hazardous wastes generated in the research and development laboratories, process operations, and maintenance and transportation functions throughout the Hanford Site. Storage operations began at the 2727-S NRDWS Facility March 14, 1983, and continued until December 30, 1986, when the last shipment of materials from the facility took place. These storage operations have been moved to the new 616 NRDWS Facility, which is an interim status unit located between the 200 East and 200 West Areas of the Hanford Site

Closure of the concrete supercontainer in hot cell under thermal load

Energy Technology Data Exchange (ETDEWEB)

Craeye, Bart, E-mail: bart.craeye@artesis.b [Artesis Univerity College of Antwerp, Applied Engineering and Technology, Antwerp (Belgium); De Schutter, Geert [Magnel Laboratory for Concrete Research, Ghent University, Technologiepark-Zwijnaarde 904, 9052 Ghent (Belgium); Wacquier, William; Van Humbeeck, Hughes [ONDRAF/NIRAS, Belgian Agency for Radioactive Waste and Enriched Fissile Materials (Belgium); Van Cotthem, Alain [Tractebel Development Engineering, Consulting Company (Belgium); Areias, Lou [SCK.CEN, Belgian Nuclear Research Center (Belgium)

2011-05-15

Research highlights: We model the behaviour of the supercontainer for the disposal of high-level waste and spent fuel assemblies during fabrication at ground surface. We study the early-age cracking behaviour of the buffer and evaluate the crack creating mechanisms. In case accurate measures are taken, cracking of the buffer can be avoided. - Abstract: For the final disposal of long-lived, heat-emitting vitrified high-level waste (HLW) in a clayey host rock, an intensive study is conducted to investigate the early-age behaviour of concrete supercontainers. Self-compacting concrete (SCC) is taken as the reference concrete type as it facilitates the casting process in combination with an improved homogeneity compared to the traditional concrete compositions. A laboratory characterization program is conducted to obtain the relevant thermal, mechanical and maturity-related properties of the SCC. These obtained data are implemented into the material database of the finite element tool HEAT to study the behaviour of the concrete layers during the different construction stages of the supercontainer: (i) Stage 1: Fabrication of the concrete buffer inside a stainless steel envelope. No early-age cracking is expected in case accurate measures are taken to reduce the thermal gradient between the outer surface and the middle of the buffer, e.g. by providing insulation and excluding wind. (ii) Stages 2-4: Emplacement of the carbon steel overpack containing the HLW canisters, filling the remaining annular gap with cementitious filler and closure by fitting the lid under thermal load. The construction stages (2-4) for the closure of the supercontainer are executed in hot cell. In this study, the crack creating mechanism and the behaviour of the concrete supercontainer during these construction stages in hot cell are investigated. In case precautionary measures are taken, such as reducing the coefficient of thermal expansion (CTE) of the overpack, prolonging the preceding cooling

Closure of the concrete supercontainer in hot cell under thermal load

International Nuclear Information System (INIS)

Craeye, Bart; De Schutter, Geert; Wacquier, William; Van Humbeeck, Hughes; Van Cotthem, Alain; Areias, Lou

2011-01-01

Research highlights: → We model the behaviour of the supercontainer for the disposal of high-level waste and spent fuel assemblies during fabrication at ground surface. → We study the early-age cracking behaviour of the buffer and evaluate the crack creating mechanisms. → In case accurate measures are taken, cracking of the buffer can be avoided. - Abstract: For the final disposal of long-lived, heat-emitting vitrified high-level waste (HLW) in a clayey host rock, an intensive study is conducted to investigate the early-age behaviour of concrete supercontainers. Self-compacting concrete (SCC) is taken as the reference concrete type as it facilitates the casting process in combination with an improved homogeneity compared to the traditional concrete compositions. A laboratory characterization program is conducted to obtain the relevant thermal, mechanical and maturity-related properties of the SCC. These obtained data are implemented into the material database of the finite element tool HEAT to study the behaviour of the concrete layers during the different construction stages of the supercontainer: (i) Stage 1: Fabrication of the concrete buffer inside a stainless steel envelope. No early-age cracking is expected in case accurate measures are taken to reduce the thermal gradient between the outer surface and the middle of the buffer, e.g. by providing insulation and excluding wind. (ii) Stages 2-4: Emplacement of the carbon steel overpack containing the HLW canisters, filling the remaining annular gap with cementitious filler and closure by fitting the lid under thermal load. The construction stages (2-4) for the closure of the supercontainer are executed in hot cell. In this study, the crack creating mechanism and the behaviour of the concrete supercontainer during these construction stages in hot cell are investigated. In case precautionary measures are taken, such as reducing the coefficient of thermal expansion (CTE) of the overpack, prolonging the

Spring 2009 Semiannual (III.H. and I.U.) Report for the HWMA/RCRA Post-Closure Permit for the INTEC Waste Calcining Facility at the INL Site

International Nuclear Information System (INIS)

Boehmer, Ann M.

2009-01-01

The Waste Calcining Facility is located at the Idaho Nuclear Technology and Engineering Center. In 1999, the Waste Calcining Facility was closed under and approved Hazardous Waste Management Act/Resource Conservation and Recovery Act Closure plan. Vessels and spaces were grouted and then covered with a concrete cap. This permit sets forth procedural requirements for groundwater characterization and monitoring, maintenance, and inspections of the Waste Calcining Facility to ensure continued protection of human health and the environment.

Spring 2009 Semiannual (III.H. and I.U.) Report for the HWMA/RCRA Post-Closure Permit for the INTEC Waste Calcining Facility at the INL Site

Energy Technology Data Exchange (ETDEWEB)

Boehmer, Ann M.

2009-05-31

The Waste Calcining Facility is located at the Idaho Nuclear Technology and Engineering Center. In 1999, the Waste Calcining Facility was closed under and approved Hazardous Waste Management Act/Resource Conservation and Recovery Act Closure plan. Vessels and spaces were grouted and then covered with a concrete cap. This permit sets forth procedural requirements for groundwater characterization and monitoring, maintenance, and inspections of the Waste Calcining Facility to ensure continued protection of human health and the environment.

CLOSURE REPORT FOR CORRECTIVE ACTION UNIT 115: AREA 25 TEST CELL A FACILITY, NEVADA TEST SITE, NEVADA

International Nuclear Information System (INIS)

2006-01-01

This Closure Report (CR) describes the activities performed to close CAU 115, Area 25 Test Cell A Facility, as presented in the NDEP-approved SAFER Plan (NNSA/NSO, 2004). The SAFER Plan includes a summary of the site history, process knowledge, and closure standards. This CR provides a summary of the completed closure activities, documentation of waste disposal, and analytical and radiological data to confirm that the remediation goals were met and to document final site conditions. The approved closure alternative as presented in the SAFER Plan for CAU 115 (NNSA/NSO, 2004) was clean closure; however, closure in place was implemented under a Record of Technical Change (ROTC) to the SAFER Plan when radiological surveys indicated that the concrete reactor pad was radiologically activated and could not be decontaminated to meet free release levels. The ROTC is included as Appendix G of this report. The objectives of closure were to remove any trapped residual liquids and gases, dispose regulated and hazardous waste, decontaminate removable radiological contamination, demolish and dispose aboveground structures, remove the dewar as a best management practice (BMP), and characterize and restrict access to all remaining radiological contamination. Radiological contaminants of concern (COCs) included cobalt-60, cesium-137, strontium-90, uranium-234/235/236/238, and plutonium-239/240. Additional COCs included Resource Conservation and Recovery Act (RCRA) metals, polychlorinated biphenyls (PCBs), and asbestos

105-DR Large Sodium Fire Facility closure activities evaluation report

International Nuclear Information System (INIS)

Adler, J.G.

1996-01-01

This report evaluates the closure activities at the 105-DR Large Sodium Fire Facility. The closure activities discussed include: the closure activities for the structures, equipment, soil, and gravel scrubber; decontamination methods; materials made available for recycling or reuse; and waste management. The evaluation compares these activities to the regulatory requirements and closure plan requirements. The report concludes that the areas identified in the closure plan can be clean closed

3718-F Alkali Metal Treatment and Storage Facility Closure Plan. Revision 1

Energy Technology Data Exchange (ETDEWEB)

None

1992-11-01

The Hanford Site, located northwest of the city of Richland, Washington, houses reactors, chemical-separation systems, and related facilities used for the production of special nuclear materials, as well as for activities associated with nuclear energy development. The 300 Area of the Hanford Site contains reactor fuel manufacturing facilities and several research and development laboratories. The 3718-F Alkali Metal Treatment and Storage Facility (3718-F Facility), located in the 300 Area, was used to store and treat alkali metal wastes. Therefore, it is subject to the regulatory requirements for the storage and treatment of dangerous wastes. Closure will be conducted pursuant to the requirements of the Washington Administrative Code (WAC) 173-303-610 (Ecology 1989) and 40 CFR 270.1. Closure also will satisfy the thermal treatment facility closure requirements of 40 CFR 265.381. This closure plan presents a description of the 3718-F Facility, the history of wastes managed, and the approach that will be followed to close the facility. Only hazardous constituents derived from 3718-F Facility operations will be addressed.

Closure of hazardous and mixed radioactive waste management units at DOE facilities

International Nuclear Information System (INIS)

1990-06-01

This is document addresses the Federal regulations governing the closure of hazardous and mixed waste units subject to Resource Conservation and Recovery Act (RCRA) requirements. It provides a brief overview of the RCRA permitting program and the extensive RCRA facility design and operating standards. It provides detailed guidance on the procedural requirements for closure and post-closure care of hazardous and mixed waste management units, including guidance on the preparation of closure and post-closure plans that must be submitted with facility permit applications. This document also provides guidance on technical activities that must be conducted both during and after closure of each of the following hazardous waste management units regulated under RCRA

Resource Conservation and Recovery Act Closure Plan for the Y-12 9409-5 Tank Storage Facility

International Nuclear Information System (INIS)

1995-02-01

This document presents information on the closure of the Y-12 9409-5 Tank Storage Facility. Topics discussed include: facility description; closure history; closure performance standard; partial closure; maximum waste inventory; closure activities; schedule; and postclosure care

Required performance to the concrete structure of the accelerator facilities

International Nuclear Information System (INIS)

Irie, Masaaki; Yoshioka, Masakazu; Miyahara, Masanobu

2006-01-01

As for the accelerator facility, there is many a thing which is constructed as underground concrete structure from viewpoint such as cover of radiation and stability of the structure. Required performance to the concrete structure of the accelerator facility is the same as the general social infrastructure, but it has been possessed the feature where target performance differs largely. As for the body sentence, expressing the difference of the performance which is required from the concrete structure of the social infrastructure and the accelerator facility, construction management of the concrete structure which it plans from order of the accelerator engineering works facility, reaches to the design, supervision and operation it is something which expresses the method of thinking. In addition, in the future of material structural analysis of the concrete which uses the neutron accelerator concerning view it showed. (author)

Work for radiation shielding concrete in large-scaled radiation facilities

International Nuclear Information System (INIS)

Konomi, Shinzo; Sato, Shoni; Otake, Takao.

1980-01-01

This paper reports the radiation shielding concrete work in the construction of radiation laboratory facilities of Electrotechnical Laboratory, a Japanese Government agency for the research and development of electronic technology. The radiation shielding walls of the facilities are made of ordinary concrete, heavy weight concrete and raw iron ore. This paper particularly relates the use of ordinary concrete which constitutes the majority of such concretes. The concrete mix was determined so as to increase its specific gravity for better shielding effect, to improve mass concrete effect and to advance good workability. The tendency of the concrete to decrease its specific gravity and the temperature variations were also made on how to place concrete to secure good shielding effect and uniform quality. (author)

United Kingdom. Development plan for the eventual closure of the UK Drigg nuclear surface low level waste disposal facility

International Nuclear Information System (INIS)

2001-01-01

The Drigg site, owned and operated by BNFL, is the UK's principal site for the disposal of low level radioactive waste. The site has operated since 1959 and receives wastes from a wide range of sources including nuclear power stations, nuclear fuel cycle facilities, isotope manufacturing sites, universities, general industry and cleanup of historically contaminated sites. Disposals until the late 1980s were solely by tipping essentially loose wastes into excavated trenches. More recently, trench disposals have been phased out in preference to emplacement of containerised, conditioned wastes in concrete vaults. The standardised wasteform consists of high force compacted (or non-compactable) waste immobilised within 20 m 3 steel overpack containers by the addition of cementitious grout. Larger items of wastes are grouted directly, in situ in the vault. The disposal trenches have been completed with an interim cap, as will the vaults when filled. It is currently estimated that sufficient capacity remains at Drigg for disposals to continue until at least 2050. Post-operations it is planned that the site will enter a phase including shut down of operational facilities, emplacement of long term site closure features including a final closure cap and then to an institutional management phase. Planning has therefore been carried out as to the strategy for eventual closure of the site. This closure strategy is also underpinned by an engineering evaluation studies programme to develop and evaluate appropriate closure measures including assessment of the long term performance of such measures. This appendix summarizes some of this work

Waste Encapsulation and Storage Facility (WESF) Interim Status Closure Plan

International Nuclear Information System (INIS)

SIMMONS, F.M.

2000-01-01

This document describes the planned activities and performance standards for closing the Waste Encapsulation and Storage Facility (WESF). WESF is located within the 225B Facility in the 200 East Area on the Hanford Facility. Although this document is prepared based on Title 40 Code of Federal Regulations (CFR), Part 265, Subpart G requirements, closure of the storage unit will comply with Washington Administrative Code (WAC) 173-303-610 regulations pursuant to Section 5.3 of the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Action Plan (Ecology et al. 1996). Because the intention is to clean close WESF, postclosure activities are not applicable to this interim status closure plan. To clean close the storage unit, it will be demonstrated that dangerous waste has not been left onsite at levels above the closure performance standard for removal and decontamination. If it is determined that clean closure is not possible or environmentally is impracticable, the interim status closure plan will be modified to address required postclosure activities. WESF stores cesium and strontium encapsulated salts. The encapsulated salts are stored in the pool cells or process cells located within 225B Facility. The dangerous waste is contained within a double containment system to preclude spills to the environment. In the unlikely event that a waste spill does occur outside the capsules, operating methods and administrative controls require that waste spills be cleaned up promptly and completely, and a notation made in the operating record. Because dangerous waste does not include source, special nuclear, and by-product material components of mixed waste, radionuclides are not within the scope of this documentation. The information on radionuclides is provided only for general knowledge

303-K Radioactive Mixed-Waste Storage Facility closure plan

International Nuclear Information System (INIS)

1991-11-01

The Hanford Site, located northwest of Richland, Washington, houses reactors chemical-separation systems, and related facilities used for the production o special nuclear materials. The 300 Area of the Hanford Site contains reactor fuel manufacturing facilities and several research and development laboratories. The 303-K Radioactive Mixed-Waste Storage Facility (303-K Facility) has been used since 1943 to store various radioactive,and dangerous process materials and wastes generated by the fuel manufacturing processes in the 300 Area. The mixed wastes are stored in US Department of Transportation (DOT)-specification containers (DOT 1988). The north end of the building was used for storage of containers of liquid waste and the outside storage areas were used for containers of solid waste. Because only the north end of the building was used, this plan does not include the southern end of the building. This closure plan presents a description of the facility, the history of materials and wastes managed, and a description of the procedures that will be followed to chose the 303-K Facility as a greater than 90-day storage facility. The strategy for closure of the 303-K Facility is presented in Chapter 6.0

Mechanisms of long-term concrete degradation in LLW disposal facilities

International Nuclear Information System (INIS)

Rogers, V.C.

1987-01-01

Most low-level waste (LLW) disposal alternatives, except shallow land burial and improved shallow land burial, involve the use of concrete as an extra barrier for containment. Because concrete is a porous-type material, its moisture retention and transport properties can be characterized with parameters that are also used to characterize the geohydrologic properties of soils. Several processes can occur with the concrete to degrade it and to increase both the movement of water and contaminants through the disposal facility. The effect of these processes must be quantified in designing and estimating the long-term performance of disposal facilities. Modeling the long-term performance of LLW disposal technologies involves, first, estimating the degradation rate of the concrete in a particular facility configuration and environmental setting; second, calculating the water flow through the facility as a function of time; third, calculating the contaminant leaching usually by diffusion or dissolution mechanisms, and then coupling the facility water and contaminant outflow to a hydrogeological and environmental uptake model for environmental releases or doses

Closure Plan for the E-Area Low-Level Waste Facility

Energy Technology Data Exchange (ETDEWEB)

Cook, J.R.

2000-10-30

A closure plan has been developed to comply with the applicable requirements of the U.S. Department of Energy Order 435.2 Manual and Guidance. The plan is organized according to the specifications of the Format and Content Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Closure Plans.

Closure Plan for the E-Area Low-Level Waste Facility

International Nuclear Information System (INIS)

Cook, J.R.

2000-01-01

A closure plan has been developed to comply with the applicable requirements of the U.S. Department of Energy Order 435.2 Manual and Guidance. The plan is organized according to the specifications of the Format and Content Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Closure Plans

111-B Metal Examination Facility Concrete Tanks Characterization Plan

International Nuclear Information System (INIS)

Encke, D.B.

1997-08-01

The 111-B Metal Examination Facility was a single-story, wood frame 'L'-shaped building built on a concrete floor slab. The facility served as a fuel failure inspection facility. Irradiated fuel pieces were stored and examined in two below grade concrete storage tanks filled with water. The tanks have been filled with grout to stabilize the contamination they contained, and overall dimensions are 5 ft 9 in. (1.5 m 22.8 cm ) wide, 9 ft 1 in. (2.7 m 2.54 cm ) deep, and 10 ft 8 in. (3.0 m 20.32 cm) long, and are estimated to weigh 39 tons. The tanks were used to store and examine failed fuel rods, using water as a radiation shield. The tanks were lined with stainless steel; however, drawings show the liner has been removed from at least one tank (south tank) and was partially filled with grout. The south tank was used to contain the Sample Storage Facility, a multi-level metal storage rack for failed nuclear fuel rods (shown in drawings H-1-2889 and -2890). Both tanks were completely grouted sometime before decontamination and demolition (D ampersand D) of the above ground facility in 1984. The 111-B Metal Examination Facility contained two concrete tanks located below floor level for storage and examination of failed fuel. The tanks were filled with concrete as part of decommissioning the facility prior to 1983 (see Appendix A for description of previous work). Funding for removal and disposal of the tanks ran out before they could be properly disposed

Soil sampling and analysis plan for the 3718-F Alkali Metal Treatment and Storage Facility closure activities

Energy Technology Data Exchange (ETDEWEB)

Sonnichsen, J.C.

1997-05-01

Amendment V.13.B.b to the approved closure plan (DOE-RL 1995a) requires that a soil sampling and analysis plan be prepared and submitted to the Washington State Department of Ecology (Ecology) for review and approval. Amendment V.13.B.c requires that a diagram of the 3718-F Alkali Metal Treatment and Storage Facility unit (the treatment, storage, and disposal [TSD] unit) boundary that is to be closed, including the maximum extent of operation, be prepared and submitted as part is of the soil sampling and analysis plan. This document describes the sampling and analysis that is to be performed in response to these requirements and amends the closure plan. Specifically, this document supersedes Section 6.2, lines 43--46, and Section 7.3.6 of the closure plan. Results from the analysis will be compared to cleanup levels identified in the closure plan. These cleanup levels will be established using residential exposure assumptions in accordance with the Model Toxics Control Act (MTCA) Cleanup Regulation (Washington Administrative Code [WAC] 173-340) as required in Amendment V.13.B.I. Results of all sampling, including the raw analytical data, a summary of analytical results, a data validation package, and a narrative summary with conclusions will be provided to Ecology as specified in Amendment V.13.B.e. The results and process used to collect and analyze the soil samples will be certified by a licensed professional engineer. These results and a certificate of closure for the balance of the TSD unit, as outlined in Chapter 7.0 of the approved closure plan (storage shed, concrete pad, burn building, scrubber, and reaction tanks), will provide the basis for a closure determination.

Soil sampling and analysis plan for the 3718-F Alkali Metal Treatment and Storage Facility closure activities

International Nuclear Information System (INIS)

Sonnichsen, J.C.

1997-01-01

Amendment V.13.B.b to the approved closure plan (DOE-RL 1995a) requires that a soil sampling and analysis plan be prepared and submitted to the Washington State Department of Ecology (Ecology) for review and approval. Amendment V.13.B.c requires that a diagram of the 3718-F Alkali Metal Treatment and Storage Facility unit (the treatment, storage, and disposal [TSD] unit) boundary that is to be closed, including the maximum extent of operation, be prepared and submitted as part is of the soil sampling and analysis plan. This document describes the sampling and analysis that is to be performed in response to these requirements and amends the closure plan. Specifically, this document supersedes Section 6.2, lines 43--46, and Section 7.3.6 of the closure plan. Results from the analysis will be compared to cleanup levels identified in the closure plan. These cleanup levels will be established using residential exposure assumptions in accordance with the Model Toxics Control Act (MTCA) Cleanup Regulation (Washington Administrative Code [WAC] 173-340) as required in Amendment V.13.B.I. Results of all sampling, including the raw analytical data, a summary of analytical results, a data validation package, and a narrative summary with conclusions will be provided to Ecology as specified in Amendment V.13.B.e. The results and process used to collect and analyze the soil samples will be certified by a licensed professional engineer. These results and a certificate of closure for the balance of the TSD unit, as outlined in Chapter 7.0 of the approved closure plan (storage shed, concrete pad, burn building, scrubber, and reaction tanks), will provide the basis for a closure determination

105-DR Large Sodium Fire Facility closure plan

International Nuclear Information System (INIS)

1993-05-01

The Hanford Site, located northwest of the city of Richland, Washington, houses reactors, chemical-separation systems, and related facilities used for the production of special nuclear materials, and activities associated with nuclear energy development. The 105-DR Large Sodium Fire Facility (LSFF), which was in operation from about 1972 to 1986, was a research laboratory that occupied the former ventilation supply room on the southwest side of the 105-DR Reactor facility. The LSFF was established to provide a means of investigating fire and safety aspects associated with large sodium or other metal alkali fires in the liquid metal fast breeder reactor (LMFBR) facilities. The 105-DR Reactor facility was designed and built in the 1950's and is located in the 100-D Area of the Hanford Site. The building housed the 105-DR defense reactor, which was shut down in 1964. The LSFF was initially used only for engineering-scale alkali metal reaction studies. In addition, the Fusion Safety Support Studies program sponsored intermediate-size safety reaction tests in the LSFF with lithium and lithium lead compounds. The facility has also been used to store and treat alkali metal waste, therefore the LSFF is subject to the regulatory requirements for the storage and treatment of dangerous waste. Closure will be conducted pursuant to the requirements of the Washington Administrative Code (WAC) 173-303-610. This closure plan presents a description of the facility, the history of waste managed, and the procedures that will be followed to close the LSFF as an Alkali Metal Treatment Facility. No future use of the LSFF is expected

Surplus Facilities and Resource Conservation and Recovery Act Closure program plan, fiscal year 1992

International Nuclear Information System (INIS)

Hughes, M.C.; Wahlen, R.K.; Winship, R.A.

1991-10-01

The Surplus Facilities and Resource Conservation and Recovery Act Closure program is responsible to US Department of Energy Field Office, Richland for the safe, cost-effective surveillance, maintenance, and decommissioning of surplus facilities at the Hanford Site. The Surplus Facilities and Resource Conservation and Recovery Act Closure program is also responsible to US Department of Energy Field Office, Richland for the program management of specific Resource Conservation and Recovery Act closures at the Hanford Site. This program plan addresses only the surplus facilities. The criteria used to evaluate each factor relative to decommissioning are based on the guidelines presented by the US Department of Energy Field Office, Richland, Environmental Restoration Division. The guidelines are consistent with the Westinghouse Hanford Company commitment to decommission Hanford Site retired facilities in the safest and most cost-effective way achievable. This document outlines the plan for managing these facilities until disposal

Procedures and techniques for closure of near surface disposal facilities for radioactive waste

International Nuclear Information System (INIS)

2001-12-01

The overall objective of this report is to provide Member States with guidance on planning and implementation of closure of near surface disposal facilities for low and intermediate level radioactive waste. The specific objectives are to review closure concepts, requirements, and components of closure systems; to discuss issues and approaches to closure, including regulatory, economic, and technical aspects; and to present major examples of closure techniques used and/or considered by Member States. Some examples of closure experience from Member States are presented in the Appendix and were indexed separately

Facility Closure Report for T-Tunnel (U12T), Area 12, Nevada Test Site, Nevada

International Nuclear Information System (INIS)

2008-01-01

This Facility Closure Report (FCR) has been prepared to document the actions taken to permanently close the remaining accessible areas of U12t-Tunnel (T-Tunnel) in Area 12 of the Nevada Test Site (NTS). The closure of T-Tunnel was a prerequisite to transfer facility ownership from the Defense Threat Reduction Agency (DTRA) to the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO). Closure of the facility was accomplished with the cooperation and concurrence of both NNSA/NSO and the Nevada Division of Environmental Protection (NDEP). The purpose of this FCR is to document that the closure of T-Tunnel complied with the closure requirements specified in the Facility Closure Plan for N- and T-Tunnels Area 12, Nevada Test Site (Appendix D) and that the facility is ready for transfer to NNSA/NSO. The Facility Closure Plan (FCP) is provided in Appendix D. T-Tunnel is located approximately 42 miles north of Mercury in Area 12 of the NTS (Figure 1). Between 1970 and 1987, T-Tunnel was used for six Nuclear Weapons Effects Tests (NWETs). The tunnel was excavated horizontally into the volcanic tuffs of Rainier Mesa. The T-Tunnel complex consists of a main access drift with two NWET containment structures, a Gas Seal Plug (GSP), and a Gas Seal Door (GSD) (Figure 2). The T-Tunnel complex was mothballed in 1993 to preserve the tunnel for resumption of testing, should it happen in the future, to stop the discharge of tunnel effluent, and to prevent unauthorized access. This was accomplished by sealing the main drift GSD

2-MW plasmajet facility thermal tests of concrete

International Nuclear Information System (INIS)

Goin, K.L.

1977-07-01

A test was made in the 2-Megawatt Plasmajet Facility to obtain experimental data relative to the thermal response of concrete to incident heat flux. 14.6 cm diameter by 8.0 cm long concrete cylinders were positioned in a supersonic flow of heated nitrogen from an arc heater. The end of the concrete cylinders impacted by the flow were subjected to heat fluxes in the range of 0.13 to 0.35 kW/cm 2 . Measurements included cold wall surface heat flux and pressure distributions, surface and indepth temperatures, ablation rates, and surface emission spectrographs. The test was part of the Sandia light water reactor safety research program and complements similar tests made in the Radiant Heat Facility at heat fluxes from 0.03 to 0.12 kW/cm 2 . A description of the tests and a tabulation of test data are included

ICPP tank farm closure study. Volume 2: Engineering design files

International Nuclear Information System (INIS)

1998-02-01

Volume 2 contains the following topical sections: Tank farm heel flushing/pH adjustment; Grouting experiments for immobilization of tank farm heel; Savannah River high level waste tank 20 closure; Tank farm closure information; Clean closure of tank farm; Remediation issues; Remote demolition techniques; Decision concerning EIS for debris treatment facility; CERCLA/RCRA issues; Area of contamination determination; Containment building of debris treatment facility; Double containment issues; Characterization costs; Packaging and disposal options for the waste resulting from the total removal of the tank farm; Take-off calculations for the total removal of soils and structures at the tank farm; Vessel off-gas systems; Jet-grouted polymer and subsurface walls; Exposure calculations for total removal of tank farm; Recommended instrumentation during retrieval operations; High level waste tank concrete encasement evaluation; Recommended heavy equipment and sizing equipment for total removal activities; Tank buoyancy constraints; Grout and concrete formulas for tank heel solidification; Tank heel pH requirements; Tank cooling water; Evaluation of conservatism of vehicle loading on vaults; Typical vault dimensions and approximately tank and vault void volumes; Radiological concerns for temporary vessel off-gas system; Flushing calculations for tank heels; Grout lift depth analysis; Decontamination solution for waste transfer piping; Grout lift determination for filling tank and vault voids; sprung structure vendor data; Grout flow properties through a 2--4 inch pipe; Tank farm load limitations; NRC low level waste grout; Project data sheet calculations; Dose rates for tank farm closure tasks; Exposure and shielding calculations for grout lines; TFF radionuclide release rates; Documentation of the clean closure of a system with listed waste discharge; and Documentation of the ORNL method of radionuclide concentrations in tanks

ICPP tank farm closure study. Volume 2: Engineering design files

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

Volume 2 contains the following topical sections: Tank farm heel flushing/pH adjustment; Grouting experiments for immobilization of tank farm heel; Savannah River high level waste tank 20 closure; Tank farm closure information; Clean closure of tank farm; Remediation issues; Remote demolition techniques; Decision concerning EIS for debris treatment facility; CERCLA/RCRA issues; Area of contamination determination; Containment building of debris treatment facility; Double containment issues; Characterization costs; Packaging and disposal options for the waste resulting from the total removal of the tank farm; Take-off calculations for the total removal of soils and structures at the tank farm; Vessel off-gas systems; Jet-grouted polymer and subsurface walls; Exposure calculations for total removal of tank farm; Recommended instrumentation during retrieval operations; High level waste tank concrete encasement evaluation; Recommended heavy equipment and sizing equipment for total removal activities; Tank buoyancy constraints; Grout and concrete formulas for tank heel solidification; Tank heel pH requirements; Tank cooling water; Evaluation of conservatism of vehicle loading on vaults; Typical vault dimensions and approximately tank and vault void volumes; Radiological concerns for temporary vessel off-gas system; Flushing calculations for tank heels; Grout lift depth analysis; Decontamination solution for waste transfer piping; Grout lift determination for filling tank and vault voids; sprung structure vendor data; Grout flow properties through a 2--4 inch pipe; Tank farm load limitations; NRC low level waste grout; Project data sheet calculations; Dose rates for tank farm closure tasks; Exposure and shielding calculations for grout lines; TFF radionuclide release rates; Documentation of the clean closure of a system with listed waste discharge; and Documentation of the ORNL method of radionuclide concentrations in tanks.

Engineering study for closure of 209E facility

Energy Technology Data Exchange (ETDEWEB)

Brevick, C.H.; Heys, W.H.; Johnson, E.D.

1997-07-07

This document is an engineering study for evaluating alternatives to determine the most cost effective closure plan for the 209E Facility, Critical Mass Laboratory. This laboratory is located in the 200 East Area of the Hanford Site and contains a Critical Assembly Room and a Mix room were criticality experiments were once performed.

Engineering study for closure of 209E facility

International Nuclear Information System (INIS)

Brevick, C.H.; Heys, W.H.; Johnson, E.D.

1997-01-01

This document is an engineering study for evaluating alternatives to determine the most cost effective closure plan for the 209E Facility, Critical Mass Laboratory. This laboratory is located in the 200 East Area of the Hanford Site and contains a Critical Assembly Room and a Mix room were criticality experiments were once performed

Financial risks of post-closure custodial care for the Barnwell radioactive waste disposal facility - 16155

International Nuclear Information System (INIS)

Baird, Robert D.; Newberry, William F.

2009-01-01

This paper reports evaluations of the adequacy of the Barnwell Extended Care Fund in light of identified risks, with the conclusion that the fund is sufficient to cover the costs and uncertainties associated with planned post-closure care of the Barnwell, South Carolina low-level radioactive waste disposal facility. It reviews background information pertinent to the facility's post-closure monitoring and maintenance and describes financial responsibility for post-closure activities. It identifies and briefly characterizes the activities planned to be conducted following facility closure and presents the midrange estimate of planned post-closure costs. The paper identifies and quantifies sources of uncertainty in activities and costs planned for post-closure care and presents 50-, 80-, and 95-percent confidence levels of planned costs. The fund is currently sufficient to cover some but not all of the costs that might be incurred as a result of unplanned events. The paper identifies, characterizes, and quantifies unplanned events, possible consequences, and probabilities of occurrence. The paper presents costs that might be incurred in responding to the unplanned initiating events and identifies levels of confidence that the fund is adequate to cover such costs. (authors)

2727-S Nonradioactive Dangerous Waste Storage Facility clean closure evaluation report

International Nuclear Information System (INIS)

Luke, S.N.

1994-01-01

This report presents the analytical results of 2727-S NRDWS facility closure verification soil sampling and compares these results to clean closure criteria. The results of this comparison will determine if clean closure of the unit is regulatorily achievable. This report also serves to notify regulators that concentrations of some analytes at the site exceed sitewide background threshold levels (DOE-RL 1993b) and/or the limits of quantitation (LOQ). This report also presents a Model Toxics Control Act Cleanup (MTCA) (WAC 173-340) regulation health-based closure standard under which the unit can clean close in lieu of closure to background levels or LOQ in accordance with WAC 173-303-610. The health-based clean closure standard will be closure to MTCA Method B residential cleanup levels. This report reconciles all analyte concentrations reported above background or LOQ to this health-based cleanup standard. Regulator acceptance of the findings presented in this report will qualify the TSD unit for clean closure in accordance with WAC 173-303-610 without further TSD unit soil sampling, or soil removal and/or decontamination. Nondetected analytes require no further evaluation

Final closure of a low level waste disposal facility

International Nuclear Information System (INIS)

Potier, J.M.

1995-01-01

The low-level radioactive waste disposal facility operated by the Agence Nationale pour la Gestion des Dechets Radioactifs near La Hague, France was opened in 1969 and is scheduled for final closure in 1996. The last waste package was received in June 1994. The total volume of disposed waste is approximately 525,000 m 3 . The site closure consists of covering the disposal structures with a multi-layer impervious cap system to prevent rainwater from infiltrating the waste isolation system. A monitoring system has been set up to verify the compliance of infiltration rates with hydraulic performance objectives (less than 10 liters per square meter and per year)

Highly durable and low permeable concrete for LLW facilities

International Nuclear Information System (INIS)

Yanagibashi, Kunio; Saito, Toshio; Odagawa, Masaro.

1997-01-01

Concrete used for LLW facilities is required to be highly durable. The authors evaluated concrete containing glycol ether derivatives and silica fume as admixtures. Compressive strength, diffusion coefficient of water, depth of accelerated carbonation, drying shrinkage, depth of chlorides penetration and resistance to freezing and thawing were investigated using concrete specimens. Compressive strength, depth of accelerated carbonation, diffusion coefficient of 137 Cs were investigated using mortar specimens before and after irradiation of gamma rays. Results showed that using glycol ether derivatives and silica fume was effective in improving the durability. (author)

The Mixed Waste Management Facility closure and expansion at the Savannah River Site

International Nuclear Information System (INIS)

Bittner, M.F.; Frye-O'Bryant, R.C.

1992-01-01

Process wastes containing radioactive and hazardous constituents have been generated throughout the operational history of the Savannah River Site. Solid wastes containing low level radionuclides were buried in Low Level Radioactive Disposal Facility (LLRWDF). Until 1986, waste containing lead and cadmium was disposed of in the Mixed Waste Management Facility (MWMF) portion of LLRWDF. Between 1986 and 1990, waste containing F-listed hazardous rags were buried. Current Resource Conservation and Recovery Act (RCRA) regulations prohibit the disposal of these hazardous wastes at nonpermitted facilities. This paper describes the closure activities for the MWMF, completed in 1990 and plans proposed for the expansion of this closure to include the LLRWDF suspect solvent rag trenches

Concrete material characterization reinforced concrete tank structure Multi-Function Waste Tank Facility

International Nuclear Information System (INIS)

Winkel, B.V.

1995-01-01

The purpose of this report is to document the Multi-Function Waste Tank Facility (MWTF) Project position on the concrete mechanical properties needed to perform design/analysis calculations for the MWTF secondary concrete structure. This report provides a position on MWTF concrete properties for the Title 1 and Title 2 calculations. The scope of the report is limited to mechanical properties and does not include the thermophysical properties of concrete needed to perform heat transfer calculations. In the 1970's, a comprehensive series of tests were performed at Construction Technology Laboratories (CTL) on two different Hanford concrete mix designs. Statistical correlations of the CTL data were later generated by Pacific Northwest Laboratories (PNL). These test results and property correlations have been utilized in various design/analysis efforts of Hanford waste tanks. However, due to changes in the concrete design mix and the lower range of MWTF operating temperatures, plus uncertainties in the CTL data and PNL correlations, it was prudent to evaluate the CTL data base and PNL correlations, relative to the MWTF application, and develop a defendable position. The CTL test program for Hanford concrete involved two different mix designs: a 3 kip/in 2 mix and a 4.5 kip/in 2 mix. The proposed 28-day design strength for the MWTF tanks is 5 kip/in 2 . In addition to this design strength difference, there are also differences between the CTL and MWTF mix design details. Also of interest, are the appropriate application of the MWTF concrete properties in performing calculations demonstrating ACI Code compliance. Mix design details and ACI Code issues are addressed in Sections 3.0 and 5.0, respectively. The CTL test program and PNL data correlations focused on a temperature range of 250 to 450 F. The temperature range of interest for the MWTF tank concrete application is 70 to 200 F

State Environmental Policy Act (SEPA) Checklist for the 105-DR Large Sodium Fire Facility Closure Plan

Energy Technology Data Exchange (ETDEWEB)

1990-09-01

The Hanford Site, located northwest of the city of Richland, Washington, houses reactors, chemical-separation systems, and related facilities used for the production of special nuclear materials, as well as for activities associated with nuclear energy development. The 105-DR Large Sodium Fire Facility (LSFF), which was in operation from about 1972 to 1986, was a research laboratory that occupied the former ventilation supply room on the southwest side of the 105-DR Reactor facility. The LSFF was established to provide means of investigating fire and safety aspects associated with large sodium or other metal alkali fires in the liquid metal fast breeder reactor (LMFBR) facilities. The 105-DR Reactor facility was designed and built in the 1950's and is located in the 100-D Area of the Hanford Site. The building housed the DR defense reactor, which was shut down in 1964. The LSFF is subject to the regulatory requirements for the storage and treatment of dangerous wastes. Clean closure is the proposed method of closure for the LSFF. Closure will be conducted pursuant to the requirements of the Washington Administrative Code (WAC) 173-303-610 (Ecology 1989). This closure plan presents a description of the facility, the history of wastes managed, and the procedures that will be followed to close the LSFF as an Alkali Metal Treatment Facility. No future use of the LSFF is expected.

Results of Washington's phase two study on closure requirements for the Hanford commercial low-level waste facility

International Nuclear Information System (INIS)

Anderson, D.C.; Hana, S.L.

1989-01-01

This paper reports on the closure design objectives and cover alternatives resulting from the state of Washington's phase two study on closure and long-term care for the Hanford commercial low-level radioactive waste disposal facility. Four approaches to dealing with subsidence and two cover design alternatives are discussed in this paper, along with information on each layer of each cover. Objectives for closure of the Hanford low-level waste facility are also discussed

105-DR Large Sodium Fire Facility closure plan. Revision 1

Energy Technology Data Exchange (ETDEWEB)

1993-05-01

The Hanford Site, located northwest of the city of Richland, Washington, houses reactors, chemical-separation systems, and related facilities used for the production of special nuclear materials, and activities associated with nuclear energy development. The 105-DR Large Sodium Fire Facility (LSFF), which was in operation from about 1972 to 1986, was a research laboratory that occupied the former ventilation supply room on the southwest side of the 105-DR Reactor facility. The LSFF was established to provide a means of investigating fire and safety aspects associated with large sodium or other metal alkali fires in the liquid metal fast breeder reactor (LMFBR) facilities. The 105-DR Reactor facility was designed and built in the 1950`s and is located in the 100-D Area of the Hanford Site. The building housed the 105-DR defense reactor, which was shut down in 1964. The LSFF was initially used only for engineering-scale alkali metal reaction studies. In addition, the Fusion Safety Support Studies program sponsored intermediate-size safety reaction tests in the LSFF with lithium and lithium lead compounds. The facility has also been used to store and treat alkali metal waste, therefore the LSFF is subject to the regulatory requirements for the storage and treatment of dangerous waste. Closure will be conducted pursuant to the requirements of the Washington Administrative Code (WAC) 173-303-610. This closure plan presents a description of the facility, the history of waste managed, and the procedures that will be followed to close the LSFF as an Alkali Metal Treatment Facility. No future use of the LSFF is expected.

Oxide-metal corium-concrete interaction test in the Vulcano facility

International Nuclear Information System (INIS)

Journeau, Ch.; Piluso, P.; Haquet, J.F.; Saretta, S.; Boccaccio, E.; Bonnet, J.M.

2007-01-01

Corium is likely to melt through the vessel and interact with the reactor pit concrete. Corium is made of a UO 2 -rich oxidic part, in which most of the decay heat is dissipated, and of a metallic part, mainly molten steel. An experiment has been set up in the Vulcano facility in which oxidic and metallic mixtures are molten in separate furnaces and poured in a concrete cavity. Induction heating is provided to the pool upper part thanks to shielding coils, so that, in case of stratification, the lighter oxidic corium-concrete mixture receives most of the power. Pre-calculations with the TOLBIAC-ICB corium-concrete interaction code based on the phase segregation model have provided valuable information for the dimensioning of this test: a thick metallic layer (>10 kg or 4 cm) has been chosen in order to obtain significant cavity ablation profiles depending on the selected heat transfer and stratification models. Stratification of the two liquid phases is predicted to occur in less than 10 minutes. In September 2006, the experiment was performed in the Vulcano facility. The corium was made of about 15 kg of steel at 1700 C and 30 kg of oxides (70% UO 2 , 16 % ZrO 2 and 14% concrete load) above 2000 C. It was poured in a limestone-rich concrete. This concrete type was selected for the first test, since the ablation is isotropic except for the initial transient, during oxidic corium-concrete interaction tests. 32 kW of induction power have been provided to the pool during the 4-hour test. The destruction of in-concrete thermocouples indicates that ablation was first mainly radial then became isotropic. This is quite similar to the ablation progression observed during previous tests with oxidic corium interacting with this type of concrete. Important 'volcanic activity' has been observed at the corium pool surface, compared to the previous oxidic corium experiments at Vulcano. (authors)

Oxide-metal corium-concrete interaction test in the Vulcano facility

Energy Technology Data Exchange (ETDEWEB)

Journeau, Ch.; Piluso, P.; Haquet, J.F.; Saretta, S.; Boccaccio, E.; Bonnet, J.M. [CEA Cadarache, Severe Accident Mastery experimental Lab. (DEN/DTN/STRI/LMA), 13 - Saint Paul lez Durance (France)

2007-07-01

Corium is likely to melt through the vessel and interact with the reactor pit concrete. Corium is made of a UO{sub 2}-rich oxidic part, in which most of the decay heat is dissipated, and of a metallic part, mainly molten steel. An experiment has been set up in the Vulcano facility in which oxidic and metallic mixtures are molten in separate furnaces and poured in a concrete cavity. Induction heating is provided to the pool upper part thanks to shielding coils, so that, in case of stratification, the lighter oxidic corium-concrete mixture receives most of the power. Pre-calculations with the TOLBIAC-ICB corium-concrete interaction code based on the phase segregation model have provided valuable information for the dimensioning of this test: a thick metallic layer (>10 kg or 4 cm) has been chosen in order to obtain significant cavity ablation profiles depending on the selected heat transfer and stratification models. Stratification of the two liquid phases is predicted to occur in less than 10 minutes. In September 2006, the experiment was performed in the Vulcano facility. The corium was made of about 15 kg of steel at 1700 C and 30 kg of oxides (70% UO{sub 2}, 16 % ZrO{sub 2} and 14% concrete load) above 2000 C. It was poured in a limestone-rich concrete. This concrete type was selected for the first test, since the ablation is isotropic except for the initial transient, during oxidic corium-concrete interaction tests. 32 kW of induction power have been provided to the pool during the 4-hour test. The destruction of in-concrete thermocouples indicates that ablation was first mainly radial then became isotropic. This is quite similar to the ablation progression observed during previous tests with oxidic corium interacting with this type of concrete. Important 'volcanic activity' has been observed at the corium pool surface, compared to the previous oxidic corium experiments at Vulcano. (authors)

RCRA corrective action and closure

International Nuclear Information System (INIS)

1995-02-01

This information brief explains how RCRA corrective action and closure processes affect one another. It examines the similarities and differences between corrective action and closure, regulators' interests in RCRA facilities undergoing closure, and how the need to perform corrective action affects the closure of DOE's permitted facilities and interim status facilities

Final closure plan for the high-explosives open burn treatment facility at Lawrence Livermore National Laboratory Experimental Test Site 300

Energy Technology Data Exchange (ETDEWEB)

Mathews, S.

1997-04-01

This document addresses the interim status closure of the HE Open Bum Treatment Facility, as detailed by Title 22, Division 4.5, Chapter 15, Article 7 of the Califonia Code of Regulations (CCR) and by Title 40, Code of Federal Regulations (CFR) Part 265, Subpart G, ``Closure and Post Closure.`` The Closure Plan (Chapter 1) and the Post- Closure Plan (Chapter 2) address the concept of long-term hazard elimination. The Closure Plan provides for capping and grading the HE Open Bum Treatment Facility and revegetating the immediate area in accordance with applicable requirements. The Closure Plan also reflects careful consideration of site location and topography, geologic and hydrologic factors, climate, cover characteristics, type and amount of wastes, and the potential for contaminant migration. The Post-Closure Plan is designed to allow LLNL to monitor the movement, if any, of pollutants from the treatment area. In addition, quarterly inspections will ensure that all surfaces of the closed facility, including the cover and diversion ditches, remain in good repair, thus precluding the potential for contaminant migration.

The technical development on recycled aggregate concrete for nuclear facility

International Nuclear Information System (INIS)

Sukekiyo, M.; Saishu, S.; Ishikura, T.; Ishigure, K.

2000-01-01

The large amount of non-radioactive concrete waste generated by decommissioning has a very big impact on the final disposal site. Therefore, NUPEC has been developing technology which recovers at a high ratio the aggregate from the dismantling concrete with a quality which can be used to construct a new nuclear power plant. The developed high-quality recycled aggregate meets the quality standards of the natural aggregate stipulated by the Japanese architectural standard specifications for nuclear power plant facilities. As a result of these experiments, it was confirmed that the recycled concrete which used this high-quality recycled aggregate had a performance equal or better than ordinary concrete which used natural aggregate. (authors)

Republic of Korea. Closure concept development for LILW disposal facility in Republic of Korea

International Nuclear Information System (INIS)

2001-01-01

Nuclear Environment Technology Institute (NETEC) of Korea Electric Power Corporation (KEPCO) is developing near surface disposal concepts for both a rock cavern type disposal facility, and a vault type facility; two types of facility are being considered to provide more options for LILW repository siting. The conceptual design for the vault type facility will be completed in 1999. As a part of conceptual design effort, a preliminary concept for the disposal facility closure has been identified

105-DR large sodium fire facility closure Plan. Revision 2

International Nuclear Information System (INIS)

Ruck, F.A. III.

1995-03-01

The 105-DR Large Sodium Fire Facility (LSFF), which was operated 1972-1986, was a research laboratory that occupied the former ventilation supply room on the SW side of the 105-DR Reactor Facility. (The 105-DR defense reactor was shut down in 1964.) LSFF was used to investigate fire and safety aspects of large sodium or other metal alkali fires in the LMFBR facilities; it was also used to store and treat alkali metal waste. This closure plan presents a description of the unit, the history of the waste managed, and the procedures that will be followed to close the LSFF as an Alkali Metal Treatment Facility. No future use of LSFF is expected. It is located within the 100-DR-2 (source) and 100-HR-3 (groundwater) operable units, which will be addressed through the RCRA facility investigation/corrective measures study process

Structural and seismic analyses of waste facility reinforced concrete storage vaults

International Nuclear Information System (INIS)

Wang, C.Y.

1995-01-01

Facility 317 of Argonne National Laboratory consists of several reinforced concrete waste storage vaults designed and constructed in the late 1940's through the early 1960's. In this paper, structural analyses of these concrete vaults subjected to various natural hazards are described, emphasizing the northwest shallow vault. The natural phenomenon hazards considered include both earthquakes and tornados. Because these vaults are deeply embedded in the soil, the SASSI (System Analysis of Soil-Structure Interaction) code was utilized for the seismic calculations. The ultimate strength method was used to analyze the reinforced concrete structures. In all studies, moment and shear strengths at critical locations of the storage vaults were evaluated. Results of the structural analyses show that almost all the waste storage vaults meet the code requirements according to ACI 349--85. These vaults also satisfy the performance goal such that confinement of hazardous materials is maintained and functioning of the facility is not interrupted

Closure Report for Corrective Action Unit 536: Area 3 Release Site, Nevada Test Site, Nevada

International Nuclear Information System (INIS)

NSTec Environmental Restoration

2007-01-01

Corrective Action Unit (CAU) 536 is located in Area 3 of the Nevada Test Site. CAU 536 is listed in the Federal Facility Agreement and Consent Order of 1996 as Area 3 Release Site, and comprises a single Corrective Action Site (CAS): (sm b ullet) CAS 03-44-02, Steam Jenny Discharge The Nevada Division of Environmental Protection (NDEP)-approved corrective action alternative for CAS 03-44-02 is clean closure. Closure activities included removing and disposing of total petroleum hydrocarbon (TPH)- and polyaromatic hydrocarbon (PAH)-impacted soil, soil impacted with plutonium (Pu)-239, and concrete pad debris. CAU 536 was closed in accordance with the NDEP-approved CAU 536 Corrective Action Plan (CAP), with minor deviations as approved by NDEP. The closure activities specified in the CAP were based on the recommendations presented in the CAU 536 Corrective Action Decision Document (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, 2004). This Closure Report documents CAU 536 closure activities. During closure activities, approximately 1,000 cubic yards (yd3) of hydrocarbon waste in the form of TPH- and PAH-impacted soil and debris, approximately 8 yd3 of Pu-239-impacted soil, and approximately 100 yd3 of concrete debris were generated, managed, and disposed of appropriately. Additionally, a previously uncharacterized, buried drum was excavated, removed, and disposed of as hydrocarbon waste as a best management practice. Waste minimization techniques, such as the utilization of laboratory analysis to characterize and classify waste streams, were employed during the performance of closure

Seismic fragility of reinforced concrete structures in nuclear facilities

International Nuclear Information System (INIS)

Gergely, P.

1985-01-01

The failure and fragility analyses of reinforced concrete structures and elements in nuclear reactor facilities within the Seismic Safety Margins Research Program (SSMRP) at the Lawrence Livermore National Laboratory are evaluated. Uncertainties in material modeling, behavior of low shear walls, and seismic risk assessment for nonlinear response receive special attention. Problems with ductility-based spectral deamplification and prediction of the stiffness of reinforced concrete walls at low stress levels are examined. It is recommended to use relatively low damping values in connection with ductility-based response reductions. The study of static nonlinear force-deflection curves is advocated for better nonlinear dynamic response predictions

Site closure and perpetual care of a low-level radioactive waste disposal facility in semi-arid climate

International Nuclear Information System (INIS)

Singh, P.N.; Breeden, K.H.; Hana, S.L.A.

1988-01-01

A study has been performed on site closure and perpetual care and maintenance requirements for the commercially operated low-level radioactive waste (LLRW) disposal facility, referred to as the Richland Facility, on the Hanford Reservation near Richland, Washington. The study included a site assessment and identification and formulation of site specific design elements for closure and perpetual care and maintenance. This paper summarizes the observations, findings and conclusions resulting from Phase I of this study. Three release mechanisms and four destructive processes are considered in the conceptual closure design process. The release mechanisms considered include subsurface liquid movement, biological transport of wastes to the surface and subsurface gas movement. The destructive processes considered are wind erosion, biological penetration or damage of cover, vegetation destroying processes and subsidence and seismic activity. The closure design elements were developed with several key principles in mind. The primary goals were to prevent intrusion into, or exposure of, the waste; to prevent or minimize release from the trenches; to provide early warning of any release that should occur; and to provide definitive information as to whether or not any observed environmental contamination actually originated from the facility

Effects of temperature on concrete cask in a dry storage facility for spent nuclear fuels

International Nuclear Information System (INIS)

Huang Weiqing; Wu Ruixian; Zheng Yukuan

2011-01-01

In the dry storage of spent nuclear fuels,concrete cask serves both as a shielding and a structural containment. The concrete in the storage facility is expected to endure the decay heat of the spent nuclear fuel during its service life. Thus, effects of the sustaining high temperature on concrete material need be evaluated for safety of the dry storage facility. In this paper, we report an experimental program aimed at investigating possible high temperature effects on properties of concrete, with emphasis on the mechanical stability, porosity,and crack-resisting ability of concrete mixes prepared using various amounts of Portland cement, fly ash, and blast furnace slag. The experimental results obtained from concrete specimens exposed to a temperature of 94 degree C for 90 days indicate that: (1) compressive strength of the concrete remains practically unchanged; (2) the ultrasonic pulse velocity, and dynamic modulus of elasticity of the concrete decrease in early stage of the high-temperature exposure,and gradually become stable with continuing exposure; (3) shrinkage of concrete mixes exhibits an increase in early stage of the exposure and does not decrease further with time; (4) concrete mixes containing pozzolanic materials,including fly ash and blast furnace slag, show better temperature-resisting characteristics than those using only Portland cement. (authors)

Risk assessment associated to possible concrete degradation of a near surface disposal facility

Directory of Open Access Journals (Sweden)

Wacquier W.

2013-07-01

Full Text Available This article outlines a risk analysis of possible concrete degradation performed in the framework of the preparation of the Safety Report of ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, for the construction and operation of a near surface disposal facility of category A waste – short-lived low and intermediate level waste – in Dessel. The main degradation mechanism considered is the carbonation of different concrete components over different periods (from the building phase up to 2000 years, which induces corrosion of the rebars. A dedicated methodology mixing risk analysis and numerical modeling of concrete carbonation has been developed to assess the critical risks of the disposal facility at different periods. According to the results obtained, risk mapping was used to assess the impact of carbonation of concrete on the different components at the different stages. The most important risk is related to an extreme situation with complete removal of the earth cover and side embankment.

Risk assessment associated to possible concrete degradation of a near surface disposal facility

Science.gov (United States)

Capra, B.; Billard, Y.; Wacquier, W.; Gens, R.

2013-07-01

This article outlines a risk analysis of possible concrete degradation performed in the framework of the preparation of the Safety Report of ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, for the construction and operation of a near surface disposal facility of category A waste - short-lived low and intermediate level waste - in Dessel. The main degradation mechanism considered is the carbonation of different concrete components over different periods (from the building phase up to 2000 years), which induces corrosion of the rebars. A dedicated methodology mixing risk analysis and numerical modeling of concrete carbonation has been developed to assess the critical risks of the disposal facility at different periods. According to the results obtained, risk mapping was used to assess the impact of carbonation of concrete on the different components at the different stages. The most important risk is related to an extreme situation with complete removal of the earth cover and side embankment.

Environmental assessment: Closure of the Waste Calcining Facility (CPP-633), Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

1996-07-01

The U.S. Department of Energy (DOE) proposes to close the Waste Calcining Facility (WCF). The WCF is a surplus DOE facility located at the Idaho Chemical Processing Plant (ICPP) on the Idaho National Engineering Laboratory (INEL). Six facility components in the WCF have been identified as Resource Conservation and Recovery Ace (RCRA)-units in the INEL RCRA Part A application. The WCF is an interim status facility. Consequently, the proposed WCF closure must comply with Idaho Rules and Standards for Hazardous Waste contained in the Idaho Administrative Procedures Act (IDAPA) Section 16.01.05. These state regulations, in addition to prescribing other requirements, incorporate by reference the federal regulations, found at 40 CFR Part 265, that prescribe the requirements for facilities granted interim status pursuant to the RCRA. The purpose of the proposed action is to reduce the risk of radioactive exposure and release of hazardous constituents and eliminate the need for extensive long-term surveillance and maintenance. DOE has determined that the closure is needed to reduce potential risks to human health and the environment, and to comply with the Idaho Hazardous Waste Management Act (HWMA) requirements

E-Area Vault Concrete Material Property And Vault Durability/Degradation Projection Recommendations

Energy Technology Data Exchange (ETDEWEB)

Phifer, M. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2014-03-11

Subsequent to the 2008 E-Area Low-Level Waste Facility (ELLWF) Performance Assessment (PA) (WSRC 2008), two additional E-Area vault concrete property testing programs have been conducted (Dixon and Phifer 2010 and SIMCO 2011a) and two additional E-Area vault concrete durability modeling projections have been made (Langton 2009 and SIMCO 2012). All the information/data from these reports has been evaluated and consolidated herein by the Savannah River National Laboratory (SRNL) at the request of Solid Waste Management (SWM) to produce E-Area vault concrete hydraulic and physical property data and vault durability/degradation projection recommendations that are adequately justified for use within associated Special Analyses (SAs) and future PA updates. The Low Activity Waste (LAW) and Intermediate Level (IL) Vaults structural degradation predictions produced by Carey 2006 and Peregoy 2006, respectively, which were used as the basis for the 2008 ELLWF PA, remain valid based upon the results of the E-Area vault concrete durability simulations reported by Langton 2009 and those reported by SIMCO 2012. Therefore revised structural degradation predictions are not required so long as the mean thickness of the closure cap overlying the vaults is no greater than that assumed within Carey 2006 and Peregoy 2006. For the LAW Vault structural degradation prediction (Carey 2006), the mean thickness of the overlying closure cap was taken as nine feet. For the IL Vault structural degradation prediction (Peregoy 2006), the mean thickness of the overlying closure cap was taken as eight feet. The mean closure cap thicknesses as described here for both E-Area Vaults will be included as a key input and assumption (I&A) in the next revision to the closure plan for the ELLWF (Phifer et al. 2009). In addition, it has been identified as new input to the PA model to be assessed in the ongoing update to the new PA Information UDQE (Flach 2013). Once the UDQE is approved, the SWM Key I

E-Area Vault Concrete Material Property And Vault Durability/Degradation Projection Recommendations

International Nuclear Information System (INIS)

Phifer, M. A.

2014-01-01

Subsequent to the 2008 E-Area Low-Level Waste Facility (ELLWF) Performance Assessment (PA) (WSRC 2008), two additional E-Area vault concrete property testing programs have been conducted (Dixon and Phifer 2010 and SIMCO 2011a) and two additional E-Area vault concrete durability modeling projections have been made (Langton 2009 and SIMCO 2012). All the information/data from these reports has been evaluated and consolidated herein by the Savannah River National Laboratory (SRNL) at the request of Solid Waste Management (SWM) to produce E-Area vault concrete hydraulic and physical property data and vault durability/degradation projection recommendations that are adequately justified for use within associated Special Analyses (SAs) and future PA updates. The Low Activity Waste (LAW) and Intermediate Level (IL) Vaults structural degradation predictions produced by Carey 2006 and Peregoy 2006, respectively, which were used as the basis for the 2008 ELLWF PA, remain valid based upon the results of the E-Area vault concrete durability simulations reported by Langton 2009 and those reported by SIMCO 2012. Therefore revised structural degradation predictions are not required so long as the mean thickness of the closure cap overlying the vaults is no greater than that assumed within Carey 2006 and Peregoy 2006. For the LAW Vault structural degradation prediction (Carey 2006), the mean thickness of the overlying closure cap was taken as nine feet. For the IL Vault structural degradation prediction (Peregoy 2006), the mean thickness of the overlying closure cap was taken as eight feet. The mean closure cap thicknesses as described here for both E-Area Vaults will be included as a key input and assumption (I and A) in the next revision to the closure plan for the ELLWF (Phifer et al. 2009). In addition, it has been identified as new input to the PA model to be assessed in the ongoing update to the new PA Information UDQE (Flach 2013). Once the UDQE is approved, the SWM Key I and

Modified electrical survey for effective leakage detection at concrete hydraulic facilities

Science.gov (United States)

Lee, Bomi; Oh, Seokhoon

2018-02-01

Three original electrode arrays for the effective leakage detection of concrete hydraulic facilities through electrical resistivity surveys are proposed: 'cross-potential', 'direct-potential' and modified tomography-like arrays. The main differences with respect to the commonly used arrays are that the current line-sources are separated from potential pole lines and floated upon the water. The potential pole lines are located directly next to the facility in order to obtain intuitive data and useful interpretations of the internal conditions of the hydraulic facility. This modified configuration of the array clearly displays the horizontal variation of the electrical field around the damaged zones of the concrete hydraulic facility, and any anomalous regions that might be found between potential poles placed across the facilities. In order to facilitate the interpretation of these modified electrical surveys, a new and creative way of presenting the measurements is also proposed and an inversion approach is provided for the modified tomography-like array. A numerical modeling and two field tests were performed to verify these new arrays and interpretation methods. The cross and direct potential array implied an ability to detect small variations of the potential field near the measurement poles. The proposed array showed the overall potential distribution across the hydraulic facility which may be used to assist in the search of trouble zones within the structure, in combination with the traditional electrical resistivity array.

ASTM STANDARD GUIDE FOR EVALUATING DISPOSAL OPTIONS FOR REUSE OF CONCRETE FROM NUCLEAR FACILITY DECOMMISSIONING

International Nuclear Information System (INIS)

Phillips, Ann Marie; Meservey, Richard H.

2003-01-01

Within the nuclear industry, many contaminated facilities that require decommissioning contain huge volumes of concrete. This concrete is generally disposed of as low-level waste at a high cost. Much of the concrete is lightly contaminated and could be reused as roadbed, fill material, or aggregate for new concrete, thus saving millions of dollars. However, because of the possibility of volumetric contamination and the lack of a method to evaluate the risks and costs of reusing concrete, reuse is rarely considered. To address this problem, Argonne National Laboratory-East (ANL-E) and the Idaho National Engineering and Environmental Laboratory teamed to write a ''concrete protocol'' to help evaluate the ramifications of reusing concrete within the U.S. Department of Energy (DOE). This document, titled the Protocol for Development of Authorized Release Limits for Concrete at U.S. Department of Energy Site (1) is based on ANL-E's previously developed scrap metal recycle protocols; on the 10-step method outlined in DOE's draft handbook, Controlling Release for Reuse or Recycle of Property Containing Residual Radioactive Material (2); and on DOE Order 4500.5, Radiation Protection of the Public and the Environment (3). The DOE concrete protocol was the basis for the ASTM Standard Guide for Evaluating Disposal Options for Concrete from Nuclear Facility Decommissioning, which was written to make the information available to a wider audience outside DOE. The resulting ASTM Standard Guide is a more concise version that can be used by the nuclear industry worldwide to evaluate the risks and costs of reusing concrete from nuclear facility decommissioning. The bulk of the ASTM Standard Guide focuses on evaluating the dose and cost for each disposal option. The user calculates these from the detailed formulas and tabulated data provided, then compares the dose and cost for each disposal option to select the best option that meets regulatory requirements. With this information

HWMA/RCRA Closure Plan for the TRA Fluorinel Dissolution Process Mockup and Gamma Facilities Waste System

International Nuclear Information System (INIS)

K. Winterholler

2007-01-01

This Hazardous Waste Management Act/Resource Conservation and Recovery Act closure plan was developed for the Test Reactor Area Fluorinel Dissolution Process Mockup and Gamma Facilities Waste System, located in Building TRA-641 at the Reactor Technology Complex (RTC), Idaho National Laboratory Site, to meet a further milestone established under the Voluntary Consent Order SITE-TANK-005 Action Plan for Tank System TRA-009. The tank system to be closed is identified as VCO-SITE-TANK-005 Tank System TRA-009. This closure plan presents the closure performance standards and methods for achieving those standards

Lessons Learned from the On-Site Disposal Facility at Fernald Closure Project

International Nuclear Information System (INIS)

Kumthekar, U.A.; Chiou, J.D.

2006-01-01

The On-Site Disposal Facility (OSDF) at the U.S. Department of Energy's (DOE) Fernald Closure Project near Cincinnati, Ohio is an engineered above-grade waste disposal facility being constructed to permanently store low level radioactive waste (LLRW) and treated mixed LLRW generated during Decommissioning and Demolition (D and D) and soil remediation performed in order to achieve the final land use goal at the site. The OSDF is engineered to store 2.93 million cubic yards of waste derived from the remediation activities. The OSDF is intended to isolate its LLRW from the environment for at least 200 years and for up to 1,000 years to the extent practicable and achievable. Construction of the OSDF started in 1997 and waste placement activities will complete by the middle of April 2006 with the final cover (cap) placement over the last open cell by the end of Spring 2006. An on-site disposal alternative is considered critical to the success of many large-scale DOE remediation projects throughout the United States. However, for various reasons this cost effective alternative is not readily available in many cases. Over the last ten years Fluor Fernald Inc. has cumulated many valuable lessons learned through the complex engineering, construction, operation, and closure processes of the OSDF. Also in the last several years representatives from other DOE sites, State agencies, as well as foreign government agencies have visited the Fernald site to look for proven experiences and practices, which may be adapted for their sites. This paper present a summary of the major issues and lessons leaned at the Fernald site related to engineering, construction, operation, and closure processes for the disposal of remediation waste. The purpose of this paper is to share lessons learned and to benefit other projects considering or operating similar on-site disposal facilities from our successful experiences. (authors)

Closure requirements

International Nuclear Information System (INIS)

Hutchinson, I.P.G.; Ellison, R.D.

1992-01-01

Closure of a waste management unit can be either permanent or temporary. Permanent closure may be due to: economic factors which make it uneconomical to mine the remaining minerals; depletion of mineral resources; physical site constraints that preclude further mining and beneficiation; environmental, regulatory or other requirements that make it uneconomical to continue to develop the resources. Temporary closure can occur for a period of several months to several years, and may be caused by factors such as: periods of high rainfall or snowfall which prevent mining and waste disposal; economic circumstances which temporarily make it uneconomical to mine the target mineral; labor problems requiring a cessation of operations for a period of time; construction activities that are required to upgrade project components such as the process facilities and waste management units; and mine or process plant failures that require extensive repairs. Permanent closure of a mine waste management unit involves the provision of durable surface containment features to protect the waters of the State in the long-term. Temporary closure may involve activities that range from ongoing maintenance of the existing facilities to the installation of several permanent closure features in order to reduce ongoing maintenance. This paper deals with the permanent closure features

Closure report for N Reactor

International Nuclear Information System (INIS)

1994-01-01

This report has been prepared to satisfy Section 3156(b) of Public Law 101-189 (Reports in Connection with Permanent Closures of Department of Energy Defense Nuclear Facilities), which requires submittal of a Closure Report to Congress by the Secretary of Energy upon the permanent cessation of production operations at a US Department of Energy (DOE) defense nuclear facility (Watkins 1991). This closure report provides: (1) A complete survey of the environmental problems at the facility; (2) Budget quality data indicating the cost of environmental restoration and other remediation and cleanup efforts at the facility; (3) A proposed cleanup schedule

Closure report for N Reactor

Energy Technology Data Exchange (ETDEWEB)

1994-01-01

This report has been prepared to satisfy Section 3156(b) of Public Law 101-189 (Reports in Connection with Permanent Closures of Department of Energy Defense Nuclear Facilities), which requires submittal of a Closure Report to Congress by the Secretary of Energy upon the permanent cessation of production operations at a US Department of Energy (DOE) defense nuclear facility (Watkins 1991). This closure report provides: (1) A complete survey of the environmental problems at the facility; (2) Budget quality data indicating the cost of environmental restoration and other remediation and cleanup efforts at the facility; (3) A proposed cleanup schedule.

Economics of a small-volume low-level radioactive waste disposal facility

International Nuclear Information System (INIS)

1993-04-01

This report was prepared by the US Department of Energy National Low-Level Waste Management Program to present the results of a life-cycle cost analysis of a low-level radioactive waste disposal facility, including all support facilities, beginning in the preoperational phase and continuing through post-closure care. The disposal technology selected for this report is earth-covered concrete vaults, which use reinforced concrete vaults constructed above grade and an earth cover constructed at the end of the operational period for permanent closure. The report develops a design, cost estimate, and schedule for the base case and eight alternative scenarios involving changes in total disposal capacity, operating life, annual disposal rate, source of financing and long-term interest rates. The purpose of this analysis of alternatives is to determine the sensitivity of cost to changes in key analytical or technical parameters, thereby evaluating the influence of a broad range of conditions. The total estimated cost of each alternative is estimated and a unit disposal charge is developed

Low-level radioactive waste disposal facility closure

International Nuclear Information System (INIS)

White, G.J.; Ferns, T.W.; Otis, M.D.; Marts, S.T.; DeHaan, M.S.; Schwaller, R.G.; White, G.J.

1990-11-01

Part I of this report describes and evaluates potential impacts associated with changes in environmental conditions on a low-level radioactive waste disposal site over a long period of time. Ecological processes are discussed and baselines are established consistent with their potential for causing a significant impact to low-level radioactive waste facility. A variety of factors that might disrupt or act on long-term predictions are evaluated including biological, chemical, and physical phenomena of both natural and anthropogenic origin. These factors are then applied to six existing, yet very different, low-level radioactive waste sites. A summary and recommendations for future site characterization and monitoring activities is given for application to potential and existing sites. Part II of this report contains guidance on the design and implementation of a performance monitoring program for low-level radioactive waste disposal facilities. A monitoring programs is described that will assess whether engineered barriers surrounding the waste are effectively isolating the waste and will continue to isolate the waste by remaining structurally stable. Monitoring techniques and instruments are discussed relative to their ability to measure (a) parameters directly related to water movement though engineered barriers, (b) parameters directly related to the structural stability of engineered barriers, and (c) parameters that characterize external or internal conditions that may cause physical changes leading to enhanced water movement or compromises in stability. Data interpretation leading to decisions concerning facility closure is discussed. 120 refs., 12 figs., 17 tabs

Low-level radioactive waste disposal facility closure

Energy Technology Data Exchange (ETDEWEB)

White, G.J.; Ferns, T.W.; Otis, M.D.; Marts, S.T.; DeHaan, M.S.; Schwaller, R.G.; White, G.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1990-11-01

Part I of this report describes and evaluates potential impacts associated with changes in environmental conditions on a low-level radioactive waste disposal site over a long period of time. Ecological processes are discussed and baselines are established consistent with their potential for causing a significant impact to low-level radioactive waste facility. A variety of factors that might disrupt or act on long-term predictions are evaluated including biological, chemical, and physical phenomena of both natural and anthropogenic origin. These factors are then applied to six existing, yet very different, low-level radioactive waste sites. A summary and recommendations for future site characterization and monitoring activities is given for application to potential and existing sites. Part II of this report contains guidance on the design and implementation of a performance monitoring program for low-level radioactive waste disposal facilities. A monitoring programs is described that will assess whether engineered barriers surrounding the waste are effectively isolating the waste and will continue to isolate the waste by remaining structurally stable. Monitoring techniques and instruments are discussed relative to their ability to measure (a) parameters directly related to water movement though engineered barriers, (b) parameters directly related to the structural stability of engineered barriers, and (c) parameters that characterize external or internal conditions that may cause physical changes leading to enhanced water movement or compromises in stability. Data interpretation leading to decisions concerning facility closure is discussed. 120 refs., 12 figs., 17 tabs.

The 4843 Alkali Metal Storage Facility Closure Plan

International Nuclear Information System (INIS)

1991-06-01

The 4843 AMSF has been used primarily to provide a centralized building to receive and store dangerous and mixed alkali metal waste, including sodium and lithium, which has been generated at the Fast Flux Test Facility and at various other Hanford Site operations that used alkali metals. Most of the dangerous and mixed alkali metal waste received consists of retired equipment from liquid sodium processes. The unit continues to store material. In general, only solid alkali metal waste that is water reactive is stored at the 4843 AMSF. The 4843 AMSF will be closed in a manner consistent with Ecology guidelines and regulations (WAC 173-303-610). The general closure procedure is detailed as follows

ICPP calcined solids storage facility closure study. Volume II: Cost estimates, planning schedules, yearly cost flowcharts, and life-cycle cost estimates

International Nuclear Information System (INIS)

1998-02-01

This document contains Volume II of the Closure Study for the Idaho Chemical Processing Plant Calcined Solids Storage Facility. This volume contains draft information on cost estimates, planning schedules, yearly cost flowcharts, and life-cycle costs for the four options described in Volume I: (1) Risk-Based Clean Closure; NRC Class C fill, (2) Risk-Based Clean Closure; Clean fill, (3) Closure to landfill Standards; NRC Class C fill, and (4) Closure to Landfill Standards; Clean fill

ICPP calcined solids storage facility closure study. Volume II: Cost estimates, planning schedules, yearly cost flowcharts, and life-cycle cost estimates

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

This document contains Volume II of the Closure Study for the Idaho Chemical Processing Plant Calcined Solids Storage Facility. This volume contains draft information on cost estimates, planning schedules, yearly cost flowcharts, and life-cycle costs for the four options described in Volume I: (1) Risk-Based Clean Closure; NRC Class C fill, (2) Risk-Based Clean Closure; Clean fill, (3) Closure to landfill Standards; NRC Class C fill, and (4) Closure to Landfill Standards; Clean fill.

Long-Term Performance of Silo Concrete in Low- and Intermediate-Level Waste (LILW) Disposal Facility

International Nuclear Information System (INIS)

Jung, Hae Ryong; Kwon, Ki Jung; Lee, Seung Hyun; Lee, Sung Bok; Jeong, Yi Yeong; Yoon, Eui Sik; Kim, Do Gyeum

2012-01-01

Concrete has been considered one of the engineered barriers in the geological disposal facility for low- and intermediate-level wastes (LILW). The concrete plays major role as structural support, groundwater infiltration barrier, and transport barrier of radionuclides dissolved from radioactive wastes. It also works as a chemical barrier due to its high pH condition. However, the performance of the concrete structure decrease over a period of time because of several physical and chemical processes. After a long period of time in the future, the concrete would lose its effectiveness as a barrier against groundwater inflow and the release of radionuclides. An subsurface environment below the frost depth should be favorable for concrete longevity as temperature and moisture variation should be minimal, significantly reducing the potential of cracking due to drying shrinkage and thermal expansion and contraction. Therefore, the concrete structures of LILW disposal facilities below groundwater table are expected to have relatively longer service life than those of near-surface or surface concrete structures. LILW in Korea is considered to be disposed of in the Wolsong LILW Disposal Center which is under construction in geological formation. 100,000 waste packages are expected to be disposed in the 6 concrete silos below EL -80m in the Wolsong LILW Disposal Center as first stage. The concrete silo has been considered the main engineered barrier which plays a role to inhibit water inflow and the release of radionuclides to the environments. Although a number of processes are responsible for the degradation of the silo concrete, it is concluded that a reinforcing steel corrosion cause the failure of the silo concrete. Therefore, a concrete silo failure time is calculated based on a corrosion initiation time which takes for chloride ions to penetrate through the concrete cover, and a corrosion propagation time. This paper aims to analyze the concrete failure time in the

Long-Term Performance of Silo Concrete in Low- and Intermediate-Level Waste (LILW) Disposal Facility

Energy Technology Data Exchange (ETDEWEB)

Jung, Hae Ryong; Kwon, Ki Jung; Lee, Seung Hyun; Lee, Sung Bok; Jeong, Yi Yeong [Korea Radioactive-waste Management Corporation, Daejeon (Korea, Republic of); Yoon, Eui Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Do Gyeum [Korea Institute of Construction Technology, Goyang (Korea, Republic of)

2012-05-15

Concrete has been considered one of the engineered barriers in the geological disposal facility for low- and intermediate-level wastes (LILW). The concrete plays major role as structural support, groundwater infiltration barrier, and transport barrier of radionuclides dissolved from radioactive wastes. It also works as a chemical barrier due to its high pH condition. However, the performance of the concrete structure decrease over a period of time because of several physical and chemical processes. After a long period of time in the future, the concrete would lose its effectiveness as a barrier against groundwater inflow and the release of radionuclides. An subsurface environment below the frost depth should be favorable for concrete longevity as temperature and moisture variation should be minimal, significantly reducing the potential of cracking due to drying shrinkage and thermal expansion and contraction. Therefore, the concrete structures of LILW disposal facilities below groundwater table are expected to have relatively longer service life than those of near-surface or surface concrete structures. LILW in Korea is considered to be disposed of in the Wolsong LILW Disposal Center which is under construction in geological formation. 100,000 waste packages are expected to be disposed in the 6 concrete silos below EL -80m in the Wolsong LILW Disposal Center as first stage. The concrete silo has been considered the main engineered barrier which plays a role to inhibit water inflow and the release of radionuclides to the environments. Although a number of processes are responsible for the degradation of the silo concrete, it is concluded that a reinforcing steel corrosion cause the failure of the silo concrete. Therefore, a concrete silo failure time is calculated based on a corrosion initiation time which takes for chloride ions to penetrate through the concrete cover, and a corrosion propagation time. This paper aims to analyze the concrete failure time in the

Reinforced concrete in the intermediable-level nuclear waste repository

International Nuclear Information System (INIS)

Duffo, Gustavo

2009-01-01

The National Atomic Energy Commission (CNEA) is responsible for developing the nuclear waste disposal management programme. This programme contemplates the design and construction of a facility for the final disposal of intermediate-level radioactive wastes. The proposed model is a near-surface monolithic repository similar to those in operation in El Cabril, Spain. The design of this type of repository is based on the use of multiple, independent and redundant barriers and the model foresees a period of 300 years of institutional post-closure control. Since the vault and cover are major components of the engineered barriers, the durability of these concrete structures is an important aspect for the facility integrity. This work presents laboratory investigations performed on the corrosion susceptibility of steel rebars embedded in two different types of high performance reinforced concretes, recently developed by the National Institute of Industrial Technology (Argentine). Concretes were made with cement with Blast Furnace Slag (CAH) and Silica Fume cement (CAH + SF). The aim of this work is to predict the service life of the intermediate level radioactive waste disposal vaults from data obtained from electrochemical techniques. Besides, the diffusion coefficients of aggressive species, such as chloride and carbon dioxide, were also determined. On the other hand, data obtained with corrosion sensors embedded in a vault prototype is also included. These sensors allow on-line measurements of several parameters related to the corrosion process such as rebar corrosion potential and corrosion current density; incoming oxygen flow that reaches the metal surface; concrete electrical resistivity; chloride concentration and internal concrete temperature. All the information obtained from both, laboratory tests and sensors will be used for the final design of the container in order to achieve a service life more or equal than the foreseen durability for this type of

Conceptual design study of a concrete canister spent-fuel storage facility

International Nuclear Information System (INIS)

Lidfors, E.D.; Tabe, T.; Johnson, H.M.

1979-01-01

This report presents a conceptual design study for the interim storage of CANDU spent fuel in concrete canisters. The canisters will be concrete flasks, which contain fuel prepackaged in double steel containment, and will be cooled by natural air convection. This is one of the methods proposed as a potential alternative to water pool storage. A preliminary study of this concept was done by CAFS (Committee Assessing Fuel Storage), and WNRE (Whiteshell Nuclear Research Establishment) is currently conducting a development and demonstration program. This study of a central facility for the storage of all Canadian spent fuel arisings to the year 2000 was completed in 1975. A brief description of the facilities required and the operations involved, a summary of costs, a survey of the monitoring requirements and a prediction of the personnel exposures associated with this method of storing spent fuel are reported here. The estimated total cost of interim storage in cylindrical canisters at a central site is $6.02/kg U (1975 dollars). Approximately half of this cost is incurred in the shipment of fuel from the reactors to the storage facility. (author)

Design and analysis of PCRV core cavity closure

International Nuclear Information System (INIS)

Lee, T.T.; Schwartz, A.A.; Koopman, D.C.A.

1980-05-01

Design requirements and considerations for a core cavity closure which led to the choice of a concrete closure with a toggle hold-down as the design for the Gas-Cooled Fast Breeder Reactor (GCFR) plant are discussed. A procedure for preliminary stress analysis of the closure by means of a three-dimensional finite element method is described. A limited parametric study using this procedure indicates the adequacy of the present closure design and the significance of radial compression developed as a result of inclined support reaction

Summary review of Mound Facility's experience in decontamination of concrete

International Nuclear Information System (INIS)

Combs, A.B.; Davis, W.P.; Garner, J.M.; Geichman, J.R.

1980-01-01

Most of the current concrete decontamination work at Mound Facility involves surfaces that are contaminated with plutonium-238. Approximately 60,000 sq. ft. of concrete floors will have to be decontaminated in Mound's current Decontamination and Decommissioning (D and D) Project. Although most of these surfaces are partially protected by a barrier (tile or paint), contaminated water and acid have penetrated these barriers. The technique for decontaminating these floors is desribed. The initial cleaning of the floor involes standard water and detergent. Acids are not used in cleaning as they tend to drive the contamination deeper into the concrete surface. Next, the floor tile is manually removed inside a temporary enclosure under negative and filtered ventilation. Finally, layers of contaminated concrete are mechanically removed inside the ventilated enclosure. The suspected depth and surface area of contamination determines the type of mechanical tool used. In summary, several generic methods of concrete decontamination can be utilized: chemical, such as water, detergent, acids, paint remover, strippable paints, etc.; rotary using sanders, grinders, scarifiers, etc.; impact such as pressure washers (hydrolasers), particle blasters, scabblers, needlers, spallers, paving and rock breakers, ram hoes, etc. The particular method used depends on several factors: surface and area involved; depth of contamination; cost and availability of equipment; usage safety and radiological control; and waste generated

Biological shielding design and qualification of concreting process for construction of electron beam irradiation facility

International Nuclear Information System (INIS)

Petwal, V.C.; Kumar, P.; Suresh, N.; Parchani, G.; Dwivedi, J.; Thakurta, A.C.

2011-01-01

A technology demonstration facility for irradiation of food and agricultural products is being set-up by RRCAT at Indore. The facility design is based on linear electron accelerator with maximum beam power of 10 kW and can be operated either in electron mode at 10 MeV or photon modes at 5/7.5 MeV. Biological shielding has been designed in accordance with NCRP 51 to achieve dose rate at all accessible points outside the irradiation vault less than the permissible limit of 0.1 mR/hr. In addition to radiation attenuation property, concrete must have satisfactory mechanical properties to meet the structural requirements. There are number of site specific variables which affect the structural, thermal and radiological properties of concrete, leading to considerable difference in actual values and design values. Hence it is essential to establish a suitable site and environmental specific process to cast the concrete and qualify the process by experimental measurement. For process qualification we have cast concrete test blocks of different thicknesses up to 3.25 m and evaluated the radiological and mechanical properties by radiometry, ultrasonic and mechanical tests. In this paper we describe the biological shielding design of the facility and analyse the results of tests carried out for qualification of the process. (author)

Measurement of 36Cl induced in shielding concrete of various accelerator facilities

International Nuclear Information System (INIS)

Bessho, K.; Matsumura, H.; Matsuhiro, T.

2003-01-01

The concentrations of 36 Cl induced in shielding concrete of the various accelerators has been measured by accelerator mass spectrometry. For three kinds of accelerator facilities, SF cyclotron (Center for Nuclear Study, the University of Tokyo), 300 MeV electron LINAC (Laboratory of Nuclear Science, Tohoku University), and 12 GeV proton synchrotron (High Energy Accelerator Research Organization), the depth profiles of 36 Cl/ 35 Cl ratios in concrete samples near the beam lines were analyzed. The depth profiles of 36 Cl/ 35 Cl are consistent with those of the radioactive concentrations of 152 Eu and 60 Co, which are formed by thermal neutron capture reactions. These results imply that 36 Cl formed in shielding concrete of these accelerators is mainly produced by thermal neutron capture of 35 Cl. The maximum 36 Cl/ 35 Cl ratio of 3x10 -8 (300 MeV electron LINAC, depth of 8 cm) corresponds to the specific radioactivity of 2x10 -3 Bq/g, which is not serious for radioactive waste management in reconstruction or decommissioning of accelerator facilities, compared with specific radioactivity of 3 H, 152 Eu and 60 Co. (author)

Standard Guide for Evaluating Disposal Options for Concrete from Nuclear Facility Decommissioning

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 This standard guide defines the process for developing a strategy for dispositioning concrete from nuclear facility decommissioning. It outlines a 10-step method to evaluate disposal options for radioactively contaminated concrete. One of the steps is to complete a detailed analysis of the cost and dose to nonradiation workers (the public); the methodology and supporting data to perform this analysis are detailed in the appendices. The resulting data can be used to balance dose and cost and select the best disposal option. These data, which establish a technical basis to apply to release the concrete, can be used in several ways: (1) to show that the release meets existing release criteria, (2) to establish a basis to request release of the concrete on a case-by-case basis, (3) to develop a basis for establishing release criteria where none exists. 1.2 This standard guide is based on the “Protocol for Development of Authorized Release Limits for Concrete at U.S. Department of Energy Sites,” (1) from ...

Post-closure safety assessment of near surface disposal facilities for disused sealed radioactive sources

International Nuclear Information System (INIS)

Lee, Seunghee; Kim, Juyoul

2017-01-01

Highlights: • Post-closure safety assessment of near surface disposal facility for DSRS was performed. • Engineered vault and rock-cavern type were considered for normal and well scenario. • 14 C, 226 Ra, 241 Am were primary nuclides contributing large portion of exposure dose. • Near surface disposal of DSRSs containing 14 C, 226 Ra and 241 Am should be restricted. - Abstract: Great attention has been recently paid to the post-closure safety assessment of low- and intermediate-level radioactive waste (LILW) disposal facility for disused sealed radioactive sources (DSRSs) around the world. Although the amount of volume of DSRSs generated from industry, medicine and research and education organization was relatively small compared with radioactive wastes from commercial nuclear power plants, some DSRSs can pose a significant hazard to human health due to their high activities and long half-lives, if not appropriately managed and disposed. In this study, post-closure safety assessment was carried out for DSRSs generated from 1991 to 2014 in Korea in order to ensure long-term safety of near surface disposal facilities. Two kinds of disposal options were considered, i.e., engineered vault type disposal facility and rock-cavern type disposal facility. Rock-cavern type disposal facility has been under operation in Gyeongju city, republic of Korea since August 2015 and engineered vault type disposal facility will be constructed until December 2020 in the vicinity of rock-cavern disposal facility. Assessment endpoint was individual dose to the member of critical group, which was modeled by GoldSim, which has been widely used as probabilistic risk analysis software based on Monte Carlo simulation in the area of safety assessment of radioactive waste facilities. In normal groundwater scenario, the maximum exposure dose was extremely low, approximately 1 × 10 −7 mSv/yr, for both disposal options and satisfied the regulatory limit of 0.1 mSv/yr. However, in the

Facility Closure Report for Tunnel U16a, Area 16, Nevada Test Site, Nevada

International Nuclear Information System (INIS)

2009-01-01

U16a is not listed in the Federal Facility Agreement and Consent Order. The closure of U16a was sponsored by the Defense Threat Reduction Agency (DTRA) and performed with the cooperation of the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office and the Nevada Division of Environmental Protection. This report documents closure of this site as identified in the DTRA Fiscal Year 2008 Statement of Work, Task 6.3. Closure activities included: (1) Removing and disposing of a shack and its contents; (2) Disposing of debris from within the shack and in the vicinity of the tunnel entrance; (3) Verifying that the tunnel is empty; (4) Welding screened covers over tunnel vent holes to limit access and allow ventilation; and (5) Constructing a full-tunnel cross-section fibercrete bulkhead to prevent access to the tunnel Field activities were conducted from July to August 2008.

Decommissioning of the Risoe Hot Cell facility

International Nuclear Information System (INIS)

Carlsen, H.

1994-06-01

Nuclear fuels have been handled and examined after irradiation by physical and chemical techniques, and radiotherapy sources, mainly 60 Co, have been produced at Risoe National Laboratory since 1964. The aims of decommissioning (during 1990-94, at IAEA Stage 2 level for reactors) were to obtain safe conditions for the remaining parts of the facility and to produce clean space areas for new projects. The facility comprises 6 concrete cells, several lead-shielded steel cells, glove boxes, shielded storage for waste, a remotely operated optical microscope, a frogman area for personnel access to the concrete cells, a decontamination facility, workshops and safety installations. All steel cells, glove boxes and the microscope were emptied and removed. The concrete cells were emptied of fissile material, scientific equipment, general tools and scrap. Decontamination should facilitate waste packing and reduce amount of waste to be stored temporarily at the Risoe waste treatment facility together with highly active waste. The concrete cells were cleaned remotely by wiping, hot spot removal, by mechanical means and vacuum cleaning. The interiors of 2 cells were decontaminated by high pressure water jetting. All master-slave manipulators and part of the contaminated ventilation system at the cells were removed. The cells are left in a non-ventilated state, connected to the atmosphere by an absolute filter. The main contaminants measured before cell closure were 60 Co, 137 Cs and alpha-emitters. Dismantling, decontamination waste disposal and received doses are described. Simple techniques involving low doses were found to be very effective. (AB)

Theoretical and experimental studies for optimization of PCRV top closures

International Nuclear Information System (INIS)

Ottosen, N.S.; Andersen, S.I.

1975-01-01

The results from the remaining part of the parameter study and the preparations for the verification of an optimized design are presented. Three models have been made in the same scale and with the same depth to span ratio α as the low LM-3 model from the first investigation, i.e. α=0.35. The model LM-5 was provided with reinforcement in the tensile zone, the upper part of the closure. This reinforcement did not influence the stresses and strains in the load carrying concrete, and the dome failed at the same pressure as in the unreinforced model LM-3. However, the closure did not disintegrate, but failed due to large overall deformations causing seal leakage. In the model LM-6, the inverted dome, which is formed at higher loads as demonstrated in LM-3, was reinforced perpendicular to the supposed middle surface. This reinforcement proved to be effective, giving the dome a higher ultimate load capacity. The LM-6 test stopped due to a circumferential crack in the flange. Finally, the unreinforced LM-7 closure was tested to failure. Apart from minor changes in the flange, LM-7 was identical to LM-3 except for the excavated upper part of the concrete, which in LM-3 formed the heavily cracked tensile zone. The ultimate load and the failure mode observed for this closure were the same as for the LM-3. The experimental results are compared to finite element calculations, in which plasticity and cracking of the concrete are taken into account, and the influence of different material models for the concrete is investigated. A unique failure criterion, which includes failure of the concrete for both tensile and compressive stresses in the same mathematical expression, is proposed. Based on the results obtained from the parameter study, a new closure design is proposed, which is optimized with respect to the requirements at service conditions and ultimate load

Hungary. Closure issues for centralized waste treatment and disposal facility in Puspokszilagy, Hungary

International Nuclear Information System (INIS)

2001-01-01

The facility was commissioned in 1976. At the time that its mission was formulated, the facility was designed to collect, transport, treat as necessary and dispose all radioactive waste originating from institutional use of radioactivity. The facility is government owned and presently operated by the Budapest branch of the State Public Health and Medical Officer Services. The disposal site is located on the ridge of a hill near Puspokszilagy village approximately 40 km Northeast of Budapest. The disposal units are located in Quaternary layers of silt and clay sequences. Annual average precipitation is approximately 650-700 mm. The facility is a typical shallow land, near surface engineered type disposal unit. There are concrete trenches and shallow wells for waste disposal purposes

Hexone Storage and Treatment Facility closure plan

International Nuclear Information System (INIS)

1992-11-01

The HSTF is a storage and treatment unit subject to the requirements for the storage and treatment of dangerous waste. Closure is being conducted under interim status and will be completed pursuant to the requirements of Washington State Department of Ecology (Ecology) Dangerous Waste Regulations, Washington Administrative Code (WAC) 173-303-610 and WAC 173-303-640. Because dangerous waste does not include the source, special nuclear, and by-product material components of mixed waste, radionuclides are not within the scope of WAC 173-303 or of this closure plan. The information on radionuclides is provided only for general knowledge where appropriate. The known hazardous/dangerous waste remaining at the site before commencing other closure activities consists of the still vessels, a tarry sludge in the storage tanks, and residual contamination in equipment, piping, filters, etc. The treatment and removal of waste at the HSTF are closure activities as defined in the Resource Conservation and Recovery Act (RCRA) of 1976 and WAC 173-303

Information collection regarding geoscientific monitoring techniques during closure of underground facility in crystalline rock

International Nuclear Information System (INIS)

Hosoya, Shinichi; Yamashita, Tadashi; Iwatsuki, Teruki; Saegusa, Hiromitsu; Onoe, Hironori; Ishibashi, Masayuki

2016-01-01

The Mizunami Underground Research Laboratory (MIU) project is being pursued by the Japan Atomic Energy Agency (JAEA) to enhance the reliability of geological disposal technologies through investigations of the deep geological environment in the crystalline host rock (granite) at Mizunami City in Gifu Prefecture, central Japan. On the occasion of the reform of the entire JAEA organization in 2014, JAEA identified the critical issues on the geoscientific research program: “Development of modelling technologies for mass transport”, “Development of drift backfilling technologies” and “Development of technologies for reducing groundwater inflow”, based on the latest results of the synthesizing R and D. The purposes of the “Development of drift backfilling technologies” are to develop closure methodology and technology, and long-term monitoring technology, and to evaluate resilience of geological environment. In order to achieve the purposes, previous information from the case example of underground facility constructed in crystalline rock in Europe has been collected in this study. In particular, the boundary conditions for the closure, geological characteristics, technical specifications, and method of monitoring have been focused. The information on the international project regarding drift closure test and development of monitoring technologies has also been collected. In addition, interviews were conducted to Finnish and Swedish specialists who have experiences involving planning, construction management, monitoring, and safety assessment for the closure to obtain the technical knowledge. Based on the collected information, concept and point of attention, which are regarding drift closure testing, and planning, execution management and monitoring on the closure of MIU, have been specified. (author)

Pre/post-closure assessment of groundwater pharmaceutical fate in a wastewater‑facility-impacted stream reach

Science.gov (United States)

Bradley, Paul M.; Barber, Larry B.; Clark, Jimmy M.; Duris, Joseph W.; Foreman, William T.; Furlong, Edward T.; Givens, Carrie E.; Hubbard, Laura E.; Hutchinson, Kasey J.; Journey, Celeste A.; Keefe, Steffanie H.; Kolpin, Dana W.

2016-01-01

Pharmaceutical contamination of contiguous groundwater is a substantial concern in wastewater-impacted streams, due to ubiquity in effluent, high aqueous mobility, designed bioactivity, and to effluent-driven hydraulic gradients. Wastewater treatment facility (WWTF) closures are rare environmental remediation events; offering unique insights into contaminant persistence, long-term wastewater impacts, and ecosystem recovery processes. The USGS conducted a combined pre/post-closure groundwater assessment adjacent to an effluent-impacted reach of Fourmile Creek, Ankeny, Iowa, USA. Higher surface-water concentrations, consistent surface-water to groundwater concentration gradients, and sustained groundwater detections tens of meters from the stream bank demonstrated the importance of WWTF effluent as the source of groundwater pharmaceuticals as well as the persistence of these contaminants under effluent-driven, pre-closure conditions. The number of analytes (110 total) detected in surface water decreased from 69 prior to closure down to 8 in the first post-closure sampling event approximately 30 d later, with a corresponding 2 order of magnitude decrease in the cumulative concentration of detected analytes. Post-closure cumulative concentrations of detected analytes were approximately 5 times higher in proximal groundwater than in surface water. About 40% of the 21 contaminants detected in a downstream groundwater transect immediately before WWTF closure exhibited rapid attenuation with estimated half-lives on the order of a few days; however, a comparable number exhibited no consistent attenuation during the year-long post-closure assessment. The results demonstrate the potential for effluent-impacted shallow groundwater systems to accumulate pharmaceutical contaminants and serve as long-term residual sources, further increasing the risk of adverse ecological effects in groundwater and the near-stream ecosystem.

Post-closure safety assessment of near surface disposal facilities for disused sealed radioactive sources

Energy Technology Data Exchange (ETDEWEB)

Lee, Seunghee; Kim, Juyoul, E-mail: gracemi@fnctech.com

2017-03-15

Highlights: • Post-closure safety assessment of near surface disposal facility for DSRS was performed. • Engineered vault and rock-cavern type were considered for normal and well scenario. • {sup 14}C, {sup 226}Ra, {sup 241}Am were primary nuclides contributing large portion of exposure dose. • Near surface disposal of DSRSs containing {sup 14}C, {sup 226}Ra and {sup 241}Am should be restricted. - Abstract: Great attention has been recently paid to the post-closure safety assessment of low- and intermediate-level radioactive waste (LILW) disposal facility for disused sealed radioactive sources (DSRSs) around the world. Although the amount of volume of DSRSs generated from industry, medicine and research and education organization was relatively small compared with radioactive wastes from commercial nuclear power plants, some DSRSs can pose a significant hazard to human health due to their high activities and long half-lives, if not appropriately managed and disposed. In this study, post-closure safety assessment was carried out for DSRSs generated from 1991 to 2014 in Korea in order to ensure long-term safety of near surface disposal facilities. Two kinds of disposal options were considered, i.e., engineered vault type disposal facility and rock-cavern type disposal facility. Rock-cavern type disposal facility has been under operation in Gyeongju city, republic of Korea since August 2015 and engineered vault type disposal facility will be constructed until December 2020 in the vicinity of rock-cavern disposal facility. Assessment endpoint was individual dose to the member of critical group, which was modeled by GoldSim, which has been widely used as probabilistic risk analysis software based on Monte Carlo simulation in the area of safety assessment of radioactive waste facilities. In normal groundwater scenario, the maximum exposure dose was extremely low, approximately 1 × 10{sup −7} mSv/yr, for both disposal options and satisfied the regulatory limit

Properties of concrete for use in near surface low-level waste disposal facilities

International Nuclear Information System (INIS)

Rogers, V.; Shuman, R.; Nielson, K.; Conner, J.

1989-01-01

The majority of alternative low-level waste disposal technologies strive to isolate the radioactive waste from the environment through the implementation of engineered man-made barriers. Of the materials used in the construction of these barriers, concrete is, by far, the most prevalent. As alternative facility designs are developed, it will be necessary to assess the features and long-term performance of the technologies. Reliable assessments will depend, in part, on an accurate understanding of the engineered barriers used in construction. Towards these ends an investigation into the properties and behavior of two types of concrete was conducted. Results are presented. Two concrete mix designs were used in the investigation. The first of these employs a Type II cement with a microsilica (silica fume) admixture. The second concrete mix uses a Type V cement with a pozzolan admixture and has approximately four percent entrained air

Study of measurement method of tritium induced in concrete of high-energy proton accelerator facilities

International Nuclear Information System (INIS)

Ohtsuka, N.; Ishihama, S.; Kunifuda, T.; Hayasaka, N.; Miura, T.

2001-01-01

Various long-loved radionuclides, 3 H, 7 Be, 22 Na, 51 Cr, 54 Mn, 56 Co, 57 Co, 60 Co, 134 Cs, 152 Eu and 154 Eu, have been produced in the shielding concrete of high energy proton accelerator facility through both nuclear spallation reactions and thermal neutron capture reactions of concrete elements, during machine operation. Tritium is the most important nuclide from the radiation protection. There were, however, few measurements of tritium concentration induced in the shielding concrete. In this study, the conditions of measurement method of tritium concentration induced in shielding concrete have been investigated using the activated shielding concrete of the 12 GeV proton beam-line tunnel at KEK and the standard rock (JG-1) irradiated of thermal neutron at the reactor. And the depth profiles of tritium induced in the shielding concrete of slow extracted proton beam line at KEK were determined using this method. (author)

Corrective Action Decision Document/Closure Report for Corrective Action Unit 559: T Tunnel Compressor/Blower Pad, Nevada Test Site

Energy Technology Data Exchange (ETDEWEB)

NSTec Environmental Restoration

2010-03-15

This Corrective Action Decision Document (CADD)/Closure Report (CR) was prepared by the Defense Threat Reduction Agency (DTRA) for Corrective Action Unit (CAU) 559, T-Tunnel Compressor/Blower Pad. This CADD/CR is consistent with the requirements of the Federal Facility Agreement and Consent Order (FFACO) agreed to by the State of Nevada, the U.S. Department of Energy, and the U.S. Department of Defense. Corrective Action Unit 559 is comprised of one Corrective Action Site (CAS): • 12-25-13, Oil Stained Soil and Concrete The purpose of this CADD/CR is to provide justification and documentation supporting the recommendation for closure in place with use restrictions for CAU 559.

Concrete sample point: 304 Concretion Facility

International Nuclear Information System (INIS)

Rollison, M.D.

1995-01-01

This report contains information concerning the analysis of concretes for volatile organic compounds. Included are the raw data for these analysis and the quality control data, the standards data, and all of the accompanying chains-of-custody records and requests for special analysis

Verification of dose rate calculation and selection study on low activation concrete in fusion facilities

International Nuclear Information System (INIS)

Oishi, Koji; Minami, Kiyoshi; Ikeda, Yujiro; Kosako, Kazuaki; Nakamura, Tomoo

1991-01-01

A concrete assembly was irradiated by D-T neutrons for 10 h, and dose rate measurement one day after shutdown has been carried out in order to provide a guide line for selection studies of low activation concrete. The experimental results were analyzed by the two dimensional calculation code DOT3.5 with its related nuclear data library GICX40 based on ENDF/B-III, however disagreement between experiment and calculation was observed in the deeper detector positions. Calculations were also performed using the nuclear data library based on ENDF/B-IV, and agreement within experimental errors was obtained at all detector positions. Selection studies for low activation concrete were performed using this nuclear data library. As a result, it was found that limestone concrete exhibited excellent properties as a low activation concrete in fusion facilities. (orig.)

Improvement of storage conditions and closure of the radioactive waste repository - Rozan

International Nuclear Information System (INIS)

Dutton, L.M.C.; Pacey, N.R.; Buckley, M.J.; Thomson, J.G.; Miller, W.; Barraclough, I.; Tomczak, W.; Mitrega, J.; Smietanski, L.

2005-01-01

The Rozan repository is a near-surface repository on the site of an ex-military fort, operated by Radioactive Waste Management Plant (RWMP). Solid or encapsulated waste is consigned to the repository. Low- and medium-activity waste produced in Poland is collected, processed, solidified and prepared for disposal at the Swierk facility. The waste is currently stored or disposed of within the fort structures, these have robust concrete walls, that provide both shielding and containment. The project, funded by the European Commission through the Phare Programme, aimed to improve the storage conditions and determine a strategy for closure achieving two key results; Stakeholder agreement to a strategy for the management and closure of the repository, and; Approval by the National Atomic Energy Agency of the safety case for the selected strategy. The strategy was selected using a multi-criteria analysis methodology at workshops that involved experts, regulators and other stakeholders. The selected strategy proposed that the waste in Facilities 3A and 8 should be left in situ and these facilities should continue to operate until the repository is closed. The waste in Rooms K7 to K9 of Facility 1 and in Facilities 2 and 3 should be retrieved, assayed, treated and packaged prior to redisposal. The short-lived waste that is retrieved from Rooms K7 to K9 of Facility 1 and Facilities 2 and 3 should be emplaced in Facility 8 subject to acceptance by the NAEA of the dose of 0.3mSv/y that might occur at long times in the future from a very unlikely scenario. When operations at the repository end, Facilities 3A and 8 should be covered with a multi-layer cap. Following selection of the strategy, assessment work was undertaken to support the production of the suite of safety cases. (author)

Resource Conservation and Recovery Act closure plan for the Intermediate-Level Transuranic Storage Facility mixed waste container storage units

International Nuclear Information System (INIS)

Nolte, E.P.; Spry, M.J.; Stanisich, S.N.

1992-11-01

This document describes the proposed plan for clean closure of the Intermediate-Level Transuranic Storage Facility mixed waste container storage units at the Idaho National Engineering Laboratory in accordance with the Resource Conservation and Recovery Act closure requirements. Descriptions of the location, size, capacity, history, and current status of the units are included. The units will be closed by removing waste containers in storage, and decontamination structures and equipment that may have contacted waste. Sufficient sampling and documentation of all activities will be performed to demonstrate clean closure. A tentative schedule is provided in the form of a milestone chart

Development of high shrinkage polyethylene terephthalate (PET) shape memory polymer tendons for concrete crack closure

Science.gov (United States)

Teall, Oliver; Pilegis, Martins; Sweeney, John; Gough, Tim; Thompson, Glen; Jefferson, Anthony; Lark, Robert; Gardner, Diane

2017-04-01

The shrinkage force exerted by restrained shape memory polymers (SMPs) can potentially be used to close cracks in structural concrete. This paper describes the physical processing and experimental work undertaken to develop high shrinkage die-drawn polyethylene terephthalate (PET) SMP tendons for use within a crack closure system. The extrusion and die-drawing procedure used to manufacture a series of PET tendon samples is described. The results from a set of restrained shrinkage tests, undertaken at differing activation temperatures, are also presented along with the mechanical properties of the most promising samples. The stress developed within the tendons is found to be related to the activation temperature, the cross-sectional area and to the draw rate used during manufacture. Comparisons with commercially-available PET strip samples used in previous research are made, demonstrating an increase in restrained shrinkage stress by a factor of two for manufactured PET filament samples.

Shield design of concrete wall between decay tank room and primary pump room in TRIGA facility

International Nuclear Information System (INIS)

Khan, M. J. H.; Rahman, M.; Haque, A.; Zulquarnain, A.; Ahmed, F. U.; Bhuiyan, S. I.

2007-01-01

The objective of this study is to recommend the radiation protection design parameters from the shielding point of view for concrete wall between the decay tank room and the primary pump room in TRIGA Mark-II research reactor facility. The shield design for this concrete wall has been performed with the help of Point-kernel Shielding Code Micro-Shield 5.05 and this design was also validated based on the measured dose rate values with Radiation Survey Meter (G-M Counter) considering the ICRP-60 (1990) recommendations for occupational dose rate limit (10 μSv/hr). The recommended shield design parameters are: (i) thickness of 114.3 cm Ilmenite-Magnetite Concrete (IMC) or 129.54 cm Ordinary Reinforced Concrete (ORC) for concrete wall A (ii) thickness of 66.04 cm Ilmenite-Magnetite Concrete (IMC) or 78.74 cm Ordinary Reinforced Concrete (ORC) for concrete wall B and (iii) door thickness of 3.175 cm Mild Steel (MS) on the entrance of decay tank room. In shielding efficiency analysis, the use of I-M concrete in the design of this concrete wall shows that it reduced the dose rate by a factor of at least 3.52 times approximately compared to ordinary reinforced concrete

Mixed Waste Management Facility (MWMF) closure, Savannah River Plant: Clay cap test section construction report

Energy Technology Data Exchange (ETDEWEB)

1988-02-26

This report contains appendices 3 through 6 for the Clay Cap Test Section Construction Report for the Mixed Waste Management Facility (MWMF) closure at the Savannah River Plant. The Clay Cap Test Program was conducted to evaluate the source, lab. permeability, in-situ permeability, and compaction characteristics, representative of kaolin clays from the Aiken, South Carolina vicinity. (KJD)

Concrete longevity overview

International Nuclear Information System (INIS)

Chang, W.; Morreale, B.

1991-01-01

A number of compact host states and unaffiliated states are currently selecting appropriate disposal technology and construction materials for their planned low-level radioactive waste (LLW) disposal facilities. Concrete is one of the candidate materials under consideration for the construction of LLW disposal facilities because of its strength, durability, abundant availability, and relatively low cost. The LLW disposal facilities must maintain intruder barrier integrity for up to 500 years, without active maintenance after the first 100 years. The ability of concrete to survive for such a long time as a construction material is a critical issue. This report provides a basic understanding of the composition and workings of concrete as a structural material in LLW disposal facilities and a description of degradation factors and state-of-the-art mitigative measures available to preserve the durability and longevity of concrete. Neither the paper nor the report is intended to be a design guidance document, and neither addresses using cement as a waste solidification agent. 5 refs., 1 tab

40 CFR 264.119 - Post-closure notices.

Science.gov (United States)

2010-07-01

...) STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES Closure... closure of each hazardous waste disposal unit, the owner or operator must submit to the local zoning... disposal unit of the facility. For hazardous wastes disposed of before January 12, 1981, the owner or...

Ultrasonic testing of a sealing construction made of salt concrete in an underground disposal facility for radioactive waste

Energy Technology Data Exchange (ETDEWEB)

Krause, Martin; Effner, Ute Antonie; Milmann, Boris; Voelker, Christoph; Wiggenhauser, Herbert [Federal Institute for Materials Research and Testing (BAM), Berlin (Germany); Mauke, Ralf [The Federal Office for Radiation Protection, Salzgitter (Germany)

2015-07-01

For the closure of radioactive waste disposal facilities engineered barriers- so called ''drift seals'' are used. The purpose of these barriers is to constrain the possible infiltration of brine and to prevent the migration of radionuclides into the biosphere. In a rock salt mine a large scale in-situ experiment of a sealing construction made of salt concrete was set up to prove the technical feasibility and operability of such barriers. In order to investigate the integrity of this structure, non-destructive ultrasonic measurements were carried out. Therefore two different methods were applied at the front side of the test-barrier: 1 Reflection measurements from boreholes 2 Ultrasonic imaging by means of scanning ultrasonic echo methods This extended abstract is a short version of an article to be published in a special edition of ASCE Journal that will briefly describe the sealing construction, the application of the non-destructive ultrasonic measurement methods and their adaptation to the onsite conditions -as well as parts of the obtained results. From this a concept for the systematic investigation of possible contribution of ultrasonic methods for quality assurance of sealing structures may be deduced.

GCFR 1/20-scale PCRV central core cavity closure model test

International Nuclear Information System (INIS)

Robinson, G.C.; Dougan, J.R.

1981-06-01

Oak Ridge National Laboratory has been conducting structural response tests of the prestressed concrete reactor vessel (PCRV) closures for the 300-MW(e) gas-cooled fast reactor demonstration power plant. This report describes the third in a series of tests of small-scale closure plug models. The model represents a redesign of the central core cavity closure plug. The primary objective was to demonstrate structural performance and ultimate load capacity of the closure plug. Secondary objectives included obtaining data on crack development and propagation and on mode of failure of the composite structure

Reclamation and closure of an oil sands tailings facility

Energy Technology Data Exchange (ETDEWEB)

Sobkowicz, J. [Thurber Engineering Ltd., Calgary, AB (Canada); Morgenstern, N. [Alberta Univ., Edmonton, AB (Canada)

2010-07-01

This PowerPoint presentation discussed methods of ensuring the successful reclamation of oil sands tailings facilities. Tailings should be reclaimed as mining proceeds in order to avoid an excessive accumulation of fluid fine tailings (FFT). The volume of mature fine tailings (MFT) in ponds should be limited in order to ensure effective tailings management. The reclaimed landforms should have good geotechnical stability and be comprised of self-sustaining native vegetation. Strength is needed to allow for timely capping and initial reclamation, and stiffness is required to minimize future settlement and to allow for the construction of a closure landscape. Reclamation strategies were presented for fines-dominated tailings; sand-depleted tailings; and sand-dominated tailings. Energy Resources Conservation Board (ERCB) criteria for tailings reclamation were discussed, and various monitoring and performance assessment strategies were presented. tabs., figs.

Molten Core - Concrete interactions in nuclear accidents. Theory and design of an experimental facility

International Nuclear Information System (INIS)

Sevon, T.

2005-11-01

In a hypothetical severe accident in a nuclear power plant, the molten core of the reactor may flow onto the concrete floor of containment building. This would cause a molten core . concrete interaction (MCCI), in which the heat transfer from the hot melt to the concrete would cause melting of the concrete. In assessing the safety of nuclear reactors, it is important to know the consequences of such an interaction. As background to the subject, this publication includes a description of the core melt stabilization concept of the European Pressurized water Reactor (EPR), which is being built in Olkiluoto in Finland. The publication includes a description of the basic theory of the interaction and the process of spalling or cracking of concrete when it is heated rapidly. A literature survey and some calculations of the physical properties of concrete and corium. concrete mixtures at high temperatures have been conducted. In addition, an equation is derived for conservative calculation of the maximum possible concrete ablation depth. The publication also includes a literature survey of experimental research on the subject of the MCCI and discussion of the results and deficiencies of the experiments. The main result of this work is the general design of an experimental facility to examine the interaction of molten metals and concrete. The main objective of the experiments is to assess the probability of spalling, or cracking, of concrete under pouring of molten material. A program of five experiments has been designed, and pre-test calculations of the experiments have been conducted with MELCOR 1.8.5 accident analysis program and conservative analytic calculations. (orig.)

ICPP calcined solids storage facility closure study. Volume III: Engineering design files

International Nuclear Information System (INIS)

1998-02-01

The following information was calculated to support cost estimates and radiation exposure calculations for closure activities at the Calcined Solids Storage Facility (CSSF). Within the estimate, volumes were calculated to determine the required amount of grout to be used during closure activities. The remaining calcine on the bin walls, supports, piping, and floor was also calculated to approximate the remaining residual calcine volumes at different stages of the removal process. The estimates for remaining calcine and vault void volume are higher than what would actually be experienced in the field, but are necessary for bounding purposes. The residual calcine in the bins may be higher than was is experienced in the field as it was assumed that the entire bin volume is full of calcine before removal activities commence. The vault void volumes are higher as the vault roof beam volumes were neglected. The estimations that follow should be considered rough order of magnitude, due to the time constraints as dictated by the project's scope of work. Should more accurate numbers be required, a new analysis would be necessary

ICPP calcined solids storage facility closure study. Volume III: Engineering design files

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

The following information was calculated to support cost estimates and radiation exposure calculations for closure activities at the Calcined Solids Storage Facility (CSSF). Within the estimate, volumes were calculated to determine the required amount of grout to be used during closure activities. The remaining calcine on the bin walls, supports, piping, and floor was also calculated to approximate the remaining residual calcine volumes at different stages of the removal process. The estimates for remaining calcine and vault void volume are higher than what would actually be experienced in the field, but are necessary for bounding purposes. The residual calcine in the bins may be higher than was is experienced in the field as it was assumed that the entire bin volume is full of calcine before removal activities commence. The vault void volumes are higher as the vault roof beam volumes were neglected. The estimations that follow should be considered rough order of magnitude, due to the time constraints as dictated by the project`s scope of work. Should more accurate numbers be required, a new analysis would be necessary.

Uranium ore mining in Spain with a focus on the closure and remediation measures in former production facilities

International Nuclear Information System (INIS)

Koch, H.; Blunck, S.; Lopez Romero, A.R.

2004-01-01

In early 2000, the uranium ore mining activities in Spain ceased. Since the middle of the last century, Spain had pushed ahead its own production of uranium concentrate with the formation of several companies (ENUSA, J.E.N.). In that period, Spain produced around 6000 t of uranium. With the completion of the operations at Andujar, La Haba and Elephante as well as Quercus at Saelices el Chico, the corporate tasks have shifted from building-up of a strategic uranium reserve to remediation and subsequent use of the locations. The operations have reached different remediation phases. While at Saelices el Chico remediation is still proceeding, the Andujar and La Haba locations are undergoing a monitoring phase as agreed for all former operating facilities. The estimated closure and remediation costs for the three operating facilities described amount to approx. 85 mio. Euro. In all three cases dealt with, however, these limited financial resources have been sufficient to successfully implement a closure and remediation concept that minimizes the risks from the facilities of uranium ore mining and processing with regard to the environment. (orig.)

An overview of technical requirements on durable concrete production for near surface disposal facilities for radioactive wastes

International Nuclear Information System (INIS)

Tolentino, Evandro; Tello, Cledola Cassia Oliveira de

2013-01-01

Radioactive waste can be generated by a wide range of activities varying from activities in hospitals to nuclear power plants, to mines and mineral processing facilities. General public have devoted nowadays considerable attention to the subject of radioactive waste management due to heightened awareness of environmental protection. The preferred strategy for the management of all radioactive waste is to contain it and to isolate it from the accessible biosphere. The Federal Government of Brazil has announced the construction for the year of 2014 and operation for the year of 2016 of a near surface disposal facility for low and intermediate level radioactive waste. The objective of this paper is to provide an overview of technical requirements related to production of durable concrete to be used in near surface disposal facilities for radioactive waste concrete structures. These requirements have been considered by researchers dealing with ongoing designing effort of the Brazilian near surface disposal facility. (author)

40 CFR 264.178 - Closure.

Science.gov (United States)

2010-07-01

... Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES Use and Management of Containers § 264.178 Closure. At closure, all hazardous waste and hazardous waste residues must be removed...

Closure Report for Corrective Action Unit 118: Area 27 Super Kukla Facility, Nevada Test Site, Nevada, Revision 0

International Nuclear Information System (INIS)

Mark Burmeister

2007-01-01

This CR provides documentation and justification for the closure of CAU 118 without further corrective action. This justification is based on process knowledge and the results of the investigative and closure activities conducted in accordance with the CAU 118 SAFER Plan: Streamlined Approach for Environmental Restoration (SAFER) Plan for CAU 118: Area 27 Super Kukla Facility, Nevada Test Site, Nevada (NNSA/NSO, 2006). The SAFER Plan provides information relating to site history as well as the scope and planning of the investigation. This CR also provides the analytical and radiological survey data to confirm that the remediation goals were met as specified in the CAU 118 SAFER Plan (NNSA/NSO, 2006). The Nevada Division of Environmental Protection (NDEP) approved the CAU 118 SAFER Plan (Murphy, 2006), which recommends closure in place with use restrictions (URs)

Installation of concrete expansion anchors at the Fast Flux Test Facility

International Nuclear Information System (INIS)

Clark, G.L.

1980-01-01

Installation criteria utilized at the Fast Flux Test Facility for concrete expansion anchors are presented. Static and dynamic load capabilities of various anchor types are discussed in relation to design loads, with particular emphasis placed on the yield load (the proportional limit). Effects of several variables (i.e., installation torque, hole diameter) are also investigated. Resolution and documentation of field problems (e.g., improper spacing, embedment, angularity) are also described. Recommendations for improving and controlling future installations are given

Evaluation of proposed panel closure modifications at WIPP

Energy Technology Data Exchange (ETDEWEB)

Allen, Lawrence E.; Silva, Matthew K.; Channell, James K.; Abel, John F.; Morgan, Dudley R.

2001-12-31

A key component in the design of the WIPP repository is the installation of concrete structures as panel seals in the intake and exhaust drifts after a panel has been filled with waste containers. As noted in the EPA final rule, the panel seal closure system is intended to block brine flow between the waste panels at the WIPP. On April 17, 2001, the DOE proposed seven modifications to the EPA concerning the design of the panel closure system. EPA approval of these modifications is necessary since the details of the panel design are specified in EPA’s final rule as a condition for WIPP certification. However, the EPA has not determined whether a rulemaking would be required for these proposed design modifications. On September 4, 2001, the DOE withdrew the request, noting that it would be resubmitted on a future date. The Environmental Evaluation Group (EEG) contracted with two engineers, Dr. John Abel and Dr. Rusty Morgan, to evaluate the proposed modifications. The EEG has accepted the conclusions and recommendations from these two experts: 1) replacement of Salado Mass Concrete with a generic salt-based concrete; 2) replacement of the explosion wall with a construction wall; 3) replacement of freshwater grouting with salt-based grouting; 4) option to allow surface or underground mixing; and 5) option to allow up to one year for completion of closure. The proposed modification to allow local carbonate river rock as aggregate is acceptable pending demonstration that no problems will exist in the resulting concrete. The proposed modification to give the contractor discretion in removal of steel forms is not supported. Instead, several recommendations are made to specifically reduce the number of forms left, thereby reducing potential migration pathways.

Seismic fragility of reinforced concrete structures and components for application to nuclear facilities

International Nuclear Information System (INIS)

Gergely, P.

1984-09-01

The failure and fragility analyses of reinforced concrete structures and elements in nuclear reactor facilities within the Seismic Safety Margins Research Program (SSMRP) at the Lawrence Livermore National Laboratory are evaluated. Uncertainties in material modeling, behavior of low shear walls, and seismic risk assessment for nonlinear response receive special attention. Problems with ductility-based spectral deamplification and prediction of the stiffness of reinforced concrete walls at low stress levels are examined. It is recommended to use relatively low damping values in connection with ductility-based response reductions. The study of static nonlinear force-deflection curves is advocated for better nonlinear dynamic response predictions. Several details of the seismic risk analysis of the Zion plant are also evaluated. 73 references

100-D Ponds closure plan. Revision 1

International Nuclear Information System (INIS)

Petersen, S.W.

1997-09-01

The 100-D Ponds is a Treatment, Storage, and Disposal (TSD) unit on the Hanford Facility that received both dangerous and nonregulated waste. This Closure Plan (Rev. 1) for the 100-D Ponds TSD unit consists of a RCRA Part A Dangerous Waste Permit Application (Rev. 3), a RCRA Closure Plan, and supporting information contained in the appendices to the plan. The closure plan consists of eight chapters containing facility description, process information, waste characteristics, and groundwater monitoring data. There are also chapters containing the closure strategy and performance standards. The strategy for the closure of the 100-D Ponds TSD unit is clean closure. Appendices A and B of the closure plan demonstrate that soil and groundwater beneath 100-D Ponds are below cleanup limits. All dangerous wastes or dangerous waste constituents or residues associated with the operation of the ponds have been removed, therefore, human health and the environment are protected. Discharges to the 100-D Ponds, which are located in the 100-DR-1 operable unit, were discontinued in June 1994. Contaminated sediment was removed from the ponds in August 1996. Subsequent sampling and analysis demonstrated that there is no contamination remaining in the ponds, therefore, this closure plan is a demonstration of clean closure

Closure Report for Corrective Action Unit 300: Surface Release Areas Nevada Test Site, Nevada

Energy Technology Data Exchange (ETDEWEB)

NSTec Environmental Restoration

2007-07-01

Corrective Action Unit (CAU) 300 is located in Areas 23, 25, and 26 of the Nevada Test Site, which is located approximately 65 miles northwest of Las Vegas, Nevada. CAU 300 is listed in the Federal Facility Agreement and Consent Order of 1996 as Surface Release Areas and is comprised of the following seven Corrective Action Sites (CASs), which are associated with the identified Building (Bldg): {sm_bullet} CAS 23-21-03, Bldg 750 Surface Discharge {sm_bullet} CAS 23-25-02, Bldg 750 Outfall {sm_bullet} CAS 23-25-03, Bldg 751 Outfall {sm_bullet} CAS 25-60-01, Bldg 3113A Outfall {sm_bullet} CAS 25-60-02, Bldg 3901 Outfall {sm_bullet} CAS 25-62-01, Bldg 3124 Contaminated Soil {sm_bullet} CAS 26-60-01, Bldg 2105 Outfall and Decon Pad The Nevada Division of Environmental Protection (NDEP)-approved corrective action alternative for CASs 23-21-03, 23-25-02, and 23-25-03 is no further action. As a best management practice, approximately 48 feet of metal piping was removed from CAS 23-25-02 and disposed of as sanitary waste. The NDEP-approved corrective action alternative for CASs 25-60-01, 25-60-02, 25-62-01, and 26-60-01, is clean closure. Closure activities for these CASs included removing and disposing of soil impacted with total petroleum hydrocarbons-diesel range organics (TPH-DRO), polychlorinated biphenyls (PCBs), semivolatile organic compounds (SVOCs), and cesium (Cs)-137, concrete impacted with TPH-DRO, and associated piping impacted with TPH-DRO. CAU 300 was closed in accordance with the NDEP-approved CAU 300 Corrective Action Plan (CAP) (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2006). The closure activities specified in the CAP were based on the recommendations presented in the CAU 300 Corrective Action Decision Document (NNSA/NSO, 2005). This Closure Report documents CAU 300 closure activities. During closure activities, approximately 40 cubic yards (yd3) of low-level waste consisting of TPH-DRO-, PCB

A study on sodium-concrete reaction

Energy Technology Data Exchange (ETDEWEB)

Bae, Jae Heum; Min, Byong Hun [Suwon University, Suwon (Korea, Republic of)

1997-07-01

A small sodium-concrete reaction facility was designed, manufactured and installed. this facility has been operated under inert gas(N{sub 2}) with different experimental variables such as sodium injection temperature, injection amount of sodium, aging period of concrete, sodium reservoir temperature. As a result, it was found that sodium injection temperature and injected amount of sodium has little effect on sodium-concrete reaction. However, sodium reservoir temperature and aging period of concrete has relatively high impact on sodium-concrete reaction. Sodium-concrete reaction model has also been developed and compared with experimental results. (Author) 51 refs., 16 tabs., 64 figs.

SU-E-T-264: New Concrete Designed and Evaluation for Megavoltage X Radiotherapy Facilities (CONTEK-RFH2).

Science.gov (United States)

Mera, M; Pereira, L; Mera, M; Pereira, L; Meilán, E; Moral, F Del; Teijeiro, A; Salgado, M; Andrade, B; Gomez, F; Fuentes-Vázquez, V; Caruncho, J; Medina, A

2012-06-01

The most common material for shielding is concrete, which can be made using various materials of different densities as aggregates. New techniques in radiotherapy, as IMRT and VMAT, require more monitor units and it is important to develop specifically designed shielding materials. Arraela S.L. has developed new concrete (CONTEK®-RFH2), which is made from an arid with a high percentage in iron (> 60%), and using the suitable sieve size, enables optimum compaction of the material and a high mass density, about 4.1-4.2 g/cm 3 . Moreover, aluminate cement, used as base, gives high resistance to high temperatures what makes this product be structurally resistant to temperatures up to 1200 ° C. The measurements were made in a LINAC Elekta SL18 to energies 6MV and 15 MV with a field size of 10×10 cm 2 for concrete samples in the form of tile 25cm×25cm with variable thickness. The linear attenuation coefficient, μm, was determined for each energy by fitting the data to Eq. 1, where Xxm is the exposure in air behind a thickness xm of the material, and X0 is the exposure in the absence of shielding. These results are compared with the ordinary concrete (2.35 g cm-3) for 6MV and 15MV energies (Ref. NCRP Report No.151). Results are tabulated in Table1. Results of attenuation are compared with ordinary concrete in Fig. 1. The new concrete CONTEK®-RFH2 increases photon attenuation and reduces the size of a shielded wall. A very high percentage in iron and a suitablesieve size approximately double the density of ordinary concrete. High mass attenuation coefficient makes this concrete an extremely desirable material for use in radiation facilities as shielding material for photon beam, and for upgrading facilities designed for less energy or less workload. © 2012 American Association of Physicists in Medicine.

Design of concrete structures important to safety of nuclear facilities

International Nuclear Information System (INIS)

2001-10-01

Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. The objective and minimum requirements for the design of civil engineering buildings/structures to be fulfilled to provide adequate assurance for safety of nuclear installations in India (such as pressurised heavy water reactor and related systems) are specified in the Safety standard for civil engineering structures important to safety of nuclear facilities. This standard is written by AERB to specify guidelines for implementation of the above civil engineering safety standard in the design of concrete structures important to safety

Precooling of concrete with flake ice

International Nuclear Information System (INIS)

Inoue, Katsuhiro; Shigenobu, Manabu; Soejima, Kenji; Noguchi, Hiroshi; Noda, Youichi; Sakaguchi, Tohru.

1989-01-01

The buildings in nuclear power stations are the reinforced concrete structures which are constructed with the massive members having much rein forcing bar quantity and relatively high strength due to the requirement of aseismatic capability, shielding and others. Also their scale is large, and in the case of a power station of one million kW class, concrete as much as 300,000 m 3 is used for one plant. Accordingly, at the time of construction, the case of stably supplying the concrete of high quality in large quantity by installing the facilities of manufacturing ready mixed concrete at construction sites is frequent. Moreover, electric power companies carry out thorough quality control to undergo the inspection before use by the Agency of Natural Resources and Energy from the aspects of materials, structures and strength. Since prestressed concrete containment vessels were adopted for No.3 and No.4 plants, the quality of concrete and the facilities for manufacturing ready mixed concrete were examined in detail. The precooling facilities for concrete and the effect of precooling are reported. (Kako, I.)

200 West Ash Pit Demolition Site closure plan

International Nuclear Information System (INIS)

1992-11-01

The Hanford Facility is owned by the US Government and operated by the US Department of Energy, Richland Field Office. Dangerous waste and mixed waste (containing both radioactive and dangerous components) are managed and produced on the Hanford Facility. Westinghouse Hanford Company is a major contractor to the US Department of Energy, Richland Field Office and serves as cooperator of the 200 West Ash Pit Demolition Site, the unit addressed in this closure plan. The 200 West Ash Pit Demolition Site Closure Plan consists of a Part A Permit Application (Revision 3) and a closure plan. An explanation of the Part A Permit Application revision is provided at the beginning of the Part A section. The closure plan consists of nine chapters and three appendices. This 200 West Ash Pit Demolition Site Closure Plan submittal contains information current as of October 15, 1992

A Concrete Framework for Environment Machines

DEFF Research Database (Denmark)

Biernacka, Malgorzata; Danvy, Olivier

2007-01-01

We materialize the common understanding that calculi with explicit substitutions provide an intermediate step between an abstract specification of substitution in the lambda-calculus and its concrete implementations. To this end, we go back to Curien’s original calculus of closures (an early...

Corrosion behaviour of steel rebars embedded in a concrete designed for the construction of an intermediate-level radioactive waste disposal facility

Directory of Open Access Journals (Sweden)

Schulz F.M.

2013-07-01

Full Text Available The National Atomic Energy Commission of the Argentine Republic is developing a nuclear waste disposal management programme that contemplates the design and construction of a facility for the final disposal of intermediate-level radioactive wastes. The repository is based on the use of multiple, independent and redundant barriers. The major components are made in reinforced concrete so, the durability of these structures is an important aspect for the facility integrity. This work presents an investigation performed on an instrumented reinforced concrete prototype specifically designed for this purpose, to study the behaviour of an intermediate level radioactive waste disposal facility from the rebar corrosion point of view. The information obtained will be used for the final design of the facility in order to guarantee a service life more or equal than the foreseen durability for this type of facilities.

Recommended management practices for operation and closure of shallow injection wells at DOE facilities

International Nuclear Information System (INIS)

1993-07-01

The Safe Drinking Water Act established the Underground Injection Control (UIC) program to ensure that underground injection of wastes does not endanger an underground source of drinking water. Under UIC regulations, an injection well is a hole in the ground, deeper than it is wide, that receives wastes or other fluid substances. Types of injection wells range from deep cased wells to shallow sumps, drywells, and drainfields. The report describes the five classes of UIC wells and summarizes relevant regulations for each class of wells and for the UIC program. The main focus of the report is Class IV and V shallow injection wells. Class IV wells are prohibited and should be closed when they are identified. Class V wells are generally authorized by rule, but EPA or a delegated state may require a permit for a Class V well. This report provides recommendations on sound operating and closure practices for shallow injection wells. In addition the report contains copies of several relevant EPA documents that provide additional information on well operation and closure. Another appendix contains information on the UIC programs in 21 states in which there are DOE facilities discharging to injection wells. The appendix includes the name of the responsible regulatory agency and contact person, a summary of differences between the state's regulations and Federal regulations, and any closure guidelines for Class IV and V wells

Evaluation of a novel trocar-site closure and comparison with a standard Carter-Thomason closure device.

Science.gov (United States)

del Junco, Michael; Okhunov, Zhamshid; Juncal, Samuel; Yoon, Renai; Landman, Jaime

2014-07-01

The aim of this study was to evaluate and compare a novel trocars-site closure device, the WECK EFx™ Endo Fascial Closure System (EFx) with the Carter-Thomason CloseSure System® (CT) for the closure of laparoscopic trocar site defects created by a 12-mm dilating trocar. We created standardized laparoscopic trocars-site abdominal wall defects in cadaver models using a standard 12-mm laparoscopic dilating trocar. Trocar defects were closed in a randomized fashion using one of the two closure systems. We recorded time and number of attempts needed for complete defect closure. In addition, we recorded the ability to maintain pneumoperitoneum, endoscopic visualization, safety, security, and facility based on the surgeon's subjective evaluations. We compared outcomes for the EFx and CT closure systems. We created 72 standardized laparoscopic trocars-site abdominal wall defects. The mean time needed for complete defect closure was 98.53 seconds (±28.9) for the EFx compared with 133.61 seconds (±54.61) for the CT (Psafety were 2.92 for EFx vs 2.19 for CT (Pvs 1.83 for EFx and CT, respectively (Pvs 2.33 for CT (P=0.022). No significant difference was observed between the EFx and the CT systems for endoscopic visualization (2.28 vs 2.50, P=0.080). In this in vitro cadaver trial, the EFx was superior in terms of time needed to complete defect closure, safety, and facility. CT was superior in terms of maintenance of pneumoperitoneum. Both systems were equal in the number of attempts needed to complete the defect closure and endoscopic visualization.

HWMA/RCRA Closure Plan for the Basin Facility Basin Water Treatment System - Voluntary Consent Order NEW-CPP-016 Action Plan

International Nuclear Information System (INIS)

Evans, S. K.

2007-01-01

This Hazardous Waste Management Act/Resource Conservation and Recovery Act closure plan for the Basin Water Treatment System located in the Basin Facility (CPP-603), Idaho Nuclear Technology and Engineering Center (INTEC), Idaho National Laboratory Site, was developed to meet future milestones established under the Voluntary Consent Order. The system to be closed includes units and associated ancillary equipment included in the Voluntary Consent Order NEW-CPP-016 Action Plan and Voluntary Consent Order SITE-TANK-005 Tank Systems INTEC-077 and INTEC-078 that were determined to have managed hazardous waste. The Basin Water Treatment System will be closed in accordance with the requirements of the Hazardous Waste Management Act/Resource Conservation and Recovery Act, as implemented by the Idaho Administrative Procedures Act 58.01.05.009 and 40 Code of Federal Regulations 265, to achieve 'clean closure' of the tank system. This closure plan presents the closure performance standards and methods of achieving those standards for the Basin Water Treatment Systems

HWMA/RCRA Closure Plan for the Basin Facility Basin Water Treatment System - Voluntary Consent Order NEW-CPP-016 Action Plan

Energy Technology Data Exchange (ETDEWEB)

Evans, S. K.

2007-11-07

This Hazardous Waste Management Act/Resource Conservation and Recovery Act closure plan for the Basin Water Treatment System located in the Basin Facility (CPP-603), Idaho Nuclear Technology and Engineering Center (INTEC), Idaho National Laboratory Site, was developed to meet future milestones established under the Voluntary Consent Order. The system to be closed includes units and associated ancillary equipment included in the Voluntary Consent Order NEW-CPP-016 Action Plan and Voluntary Consent Order SITE-TANK-005 Tank Systems INTEC-077 and INTEC-078 that were determined to have managed hazardous waste. The Basin Water Treatment System will be closed in accordance with the requirements of the Hazardous Waste Management Act/Resource Conservation and Recovery Act, as implemented by the Idaho Administrative Procedures Act 58.01.05.009 and 40 Code of Federal Regulations 265, to achieve "clean closure" of the tank system. This closure plan presents the closure performance standards and methods of achieving those standards for the Basin Water Treatment Systems.

Closure and Sealing Design Calculation

International Nuclear Information System (INIS)

T. Lahnalampi; J. Case

2005-01-01

The purpose of the ''Closure and Sealing Design Calculation'' is to illustrate closure and sealing methods for sealing shafts, ramps, and identify boreholes that require sealing in order to limit the potential of water infiltration. In addition, this calculation will provide a description of the magma that can reduce the consequences of an igneous event intersecting the repository. This calculation will also include a listing of the project requirements related to closure and sealing. The scope of this calculation is to: summarize applicable project requirements and codes relating to backfilling nonemplacement openings, removal of uncommitted materials from the subsurface, installation of drip shields, and erecting monuments; compile an inventory of boreholes that are found in the area of the subsurface repository; describe the magma bulkhead feature and location; and include figures for the proposed shaft and ramp seals. The objective of this calculation is to: categorize the boreholes for sealing by depth and proximity to the subsurface repository; develop drawing figures which show the location and geometry for the magma bulkhead; include the shaft seal figures and a proposed construction sequence; and include the ramp seal figure and a proposed construction sequence. The intent of this closure and sealing calculation is to support the License Application by providing a description of the closure and sealing methods for the Safety Analysis Report. The closure and sealing calculation will also provide input for Post Closure Activities by describing the location of the magma bulkhead. This calculation is limited to describing the final configuration of the sealing and backfill systems for the underground area. The methods and procedures used to place the backfill and remove uncommitted materials (such as concrete) from the repository and detailed design of the magma bulkhead will be the subject of separate analyses or calculations. Post-closure monitoring will not

Measurement of concentrations of {gamma}-ray emitters induced in the concrete shield of the JAERI electron linac facility

Energy Technology Data Exchange (ETDEWEB)

Endo, Akira; Kawasaki, Katsuya; Kikuchi, Masamitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Harada, Yasunori

1997-07-01

Measurement has been made to study distributions of {gamma}-ray emitters induced in the concrete shield of the JAERI electron linac facility. Core boring was carried out at seven positions to take samples from the concrete shield, and {gamma}-ray counting rates and {gamma}-ray spectra of these samples were measured with a NaI(Tl) detector and a Ge semiconductor detector, respectively. The following radionuclides were detected in the concrete samples: {sup 60}Co, {sup 134}Cs, {sup 152}Eu and {sup 154}Eu generated through thermal neutron capture reaction, and {sup 22}Na and {sup 54}Mn generated through nuclear reactions by bremsstrahlung and fast neutrons. The relation between the distributions of {gamma}-ray emitters, as a function of the depth of concrete, and the positions of core boring is discussed. (author)

Closure Report for Corrective Action Unit 240: Area 25 Vehicle Washdown Nevada Test Site, Nevada

International Nuclear Information System (INIS)

Gustafason, D.L.

2001-01-01

The Area 25 Vehicle Washdown, Corrective Action Unit (CAU) 240, was clean-closed following the approved Corrective Action Decision Document closure alternative and in accordance with the Federal Facility Agreement and Consent Order (FFACO, 1996). The CAU consists of thee Corrective Action Sites (CASs): 25-07-01 - Vehicle Washdown Area (Propellant Pad); 25-07-02 - Vehicle Washdown Area (F and J Roads Pad); and 25-07-03 - Vehicle Washdown Station (RADSAFE Pad). Characterization activities indicated that only CAS 25-07-02 (F and J Roads Pad) contained constituents of concern (COCs) above action levels and required remediation. The COCs detected were Total Petroleum Hydrocarbons (TPH) as diesel, cesium-137, and strontium-90. The F and J Roads Pad may have been used for the decontamination of vehicles and possibly disassembled engine and reactor parts from Test Cell C. Activities occurred there during the 1960s through early 1970s. The F and J Roads Pad consisted of a 9- by 5-meter (m) (30- by 15-foot [ft]) concrete pad and a 14- by 13-m (46-by 43-ft) gravel sump. The clean-closure corrective action consisted of excavation, disposal, verification sampling, backfilling, and regrading. Closure activities began on August 21, 2000, and ended on September 19, 2000. Waste disposal activities were completed on December 12, 2000. A total of 172 cubic meters (223 cubic yards) of impacted soil was excavated and disposed. The concrete pad was also removed and disposed. Verification samples were collected from the bottom and sidewalls of the excavation and analyzed for TPH diesel and 20-minute gamma spectroscopy. The sample results indicated that all impacted soil above remediation standards was removed. The closure was completed following the approved Corrective Action Plan. All impacted waste was disposed in the Area 6 Hydrocarbon Landfill. All non-impacted debris was disposed in the Area 9 Construction Landfill and the Area 23 Sanitary Landfill

Life Cycle Cost Analysis of Ready Mix Concrete Plant

Science.gov (United States)

Topkar, V. M.; Duggar, A. R.; Kumar, A.; Bonde, P. P.; Girwalkar, R. S.; Gade, S. B.

2013-11-01

India, being a developing nation is experiencing major growth in its infrastructural sector. Concrete is the major component in construction. The requirement of good quality of concrete in large quantities can be fulfilled by ready mix concrete batching and mixing plants. The paper presents a technique of applying the value engineering tool life cycle cost analysis to a ready mix concrete plant. This will help an investor or an organization to take investment decisions regarding a ready mix concrete facility. No economic alternatives are compared in this study. A cost breakdown structure is prepared for the ready mix concrete plant. A market survey has been conducted to collect realistic costs for the ready mix concrete facility. The study establishes the cash flow for the ready mix concrete facility helpful in investment and capital generation related decisions. Transit mixers form an important component of the facility and are included in the calculations. A fleet size for transit mixers has been assumed for this purpose. The life cycle cost has been calculated for the system of the ready mix concrete plant and transit mixers.

Closure Report for Corrective Action Unit 300: Surface Release Areas Nevada Test Site, Nevada

International Nuclear Information System (INIS)

NSTec Environmental Restoration

2007-01-01

Corrective Action Unit (CAU) 300 is located in Areas 23, 25, and 26 of the Nevada Test Site, which is located approximately 65 miles northwest of Las Vegas, Nevada. CAU 300 is listed in the Federal Facility Agreement and Consent Order of 1996 as Surface Release Areas and is comprised of the following seven Corrective Action Sites (CASs), which are associated with the identified Building (Bldg): (sm b ullet) CAS 23-21-03, Bldg 750 Surface Discharge (sm b ullet) CAS 23-25-02, Bldg 750 Outfall (sm b ullet) CAS 23-25-03, Bldg 751 Outfall (sm b ullet) CAS 25-60-01, Bldg 3113A Outfall (sm b ullet) CAS 25-60-02, Bldg 3901 Outfall (sm b ullet) CAS 25-62-01, Bldg 3124 Contaminated Soil (sm b ullet) CAS 26-60-01, Bldg 2105 Outfall and Decon Pad The Nevada Division of Environmental Protection (NDEP)-approved corrective action alternative for CASs 23-21-03, 23-25-02, and 23-25-03 is no further action. As a best management practice, approximately 48 feet of metal piping was removed from CAS 23-25-02 and disposed of as sanitary waste. The NDEP-approved corrective action alternative for CASs 25-60-01, 25-60-02, 25-62-01, and 26-60-01, is clean closure. Closure activities for these CASs included removing and disposing of soil impacted with total petroleum hydrocarbons-diesel range organics (TPH-DRO), polychlorinated biphenyls (PCBs), semivolatile organic compounds (SVOCs), and cesium (Cs)-137, concrete impacted with TPH-DRO, and associated piping impacted with TPH-DRO. CAU 300 was closed in accordance with the NDEP-approved CAU 300 Corrective Action Plan (CAP) (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2006). The closure activities specified in the CAP were based on the recommendations presented in the CAU 300 Corrective Action Decision Document (NNSA/NSO, 2005). This Closure Report documents CAU 300 closure activities. During closure activities, approximately 40 cubic yards (yd3) of low-level waste consisting of TPH

Measurements of 36Cl production rates from Cl, K, and Ca in concrete at the 500-MeV neutron irradiation facility of KENS

International Nuclear Information System (INIS)

Aze, T.; Fujimura, M.; Matsumura, H.; Masumoto, K.; Nakao, N.; Kawai, M.; Matsuzaki, H.; Nagai, H.

2005-01-01

In high-energy accelerator facilities, concrete components around beam lines are exposed to secondary neutrons having various energies during machine operation. The neutrons produce the various long half-life radionuclides, such as 3 H, 36 Cl, 60 Co, and 152 Eu, in the concrete. Most of the nuclides mainly produced by thermal neutron-capture reactions and their specific activities are important from the viewpoint of accelerator clearance. In previous work, the specific activities of the 36 Cl in the concretes at the various accelerator facilities have been measured and it was suggested that the 36 Cl in the concrete is useful as an indicator for thermal neutron fluence because of a characteristic of very long half life (301 kyr). However, in the concretes of the accelerator facilities over several hundreds of MeV, the 36 Cl are considerably produced by spallation from other concrete components, such as K and Ca, in addition to the thermal neutron capture of 35 Cl. The contribution of the 36 Cl productions from the spallation is unclear due to the lack of the cross sections for the neutron-induced reactions. In this work, therefore, we measured the 36 Cl production rates in concrete from Cl, K, and Ca targets in irradiation with secondary neutrons, which were produced by a bombardment of primary 500-MeV protons with W targets, at high-energy neutron-irradiation course of KENS. Samples of NaCl, K2CO 3 , and CaCO 3 were set into 7. irradiation spaces located on the depth raging from O to 320 cm from the concrete surface and irradiated for approximately one week. After the irradiation, separations of Cl from the samples were carried out radiochemically and the production rates of 36 Cl were determined by the AMS. The production rates from Cl, K, and Ca exponentially decreased with an increase of the depth from the concrete surface, and the profiles were very similar each other. Although the production rates from Cl were two orders higher than those from Ca in the same

324 Facility special-case waste assessment in support of 324 closure (TPA milestone M-89-05)

International Nuclear Information System (INIS)

Hobart, R.L.

1998-01-01

Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement Milestone M-89-05, requires US Department of Energy, Richland Operations Office to complete a 324 Facility Special-Case Waste Assessment in Support of 324 Closure. This document, HNF-1270, has been prepared with the intent of meeting this regulatory commitment. Alternatives for the special-case wastes located in the 324 Building were defined and analyzed. Based on the criteria of safety, environmental, complexity of interfaces, risk, cost, schedule, and long-term operability and maintainability, the best alternative was chosen. Waste packaging and transportation options are also included in the recommendations. The waste disposition recommendations for the B-Cell dispersibles/tank heels and High-Level Vault packaged residuals are to direct them to the Plutonium Uranium Extraction Facility (PUREX) Number 2 storage tunnel

Characterization of basin concrete in support of structural integrity demonstration for extended storage

International Nuclear Information System (INIS)

Duncan, A.

2014-01-01

Concrete core samples from C basin were characterized through material testing and analysis to verify the design inputs for structural analysis of the L Basin and to evaluate the type and extent of changes in the material condition of the concrete under extended service for fuel storage. To avoid the impact on operations, core samples were not collected from L area, but rather, several concrete core samples were taken from the C Basin prior to its closure. C basin was selected due to its similar environmental exposure and service history compared to L Basin. The microstructure and chemical composition of the concrete exposed to the water was profiled from the water surface into the wall to evaluate the impact and extent of exposure. No significant leaching of concrete components was observed. Ingress of carbonation or deleterious species was determined to be insignificant. No evidence of alkali-silica reactions (ASR) was observed. Ettringite was observed to form throughout the structure (in air voids or pores); however, the sulfur content was measured to be consistent with the initial concrete that was used to construct the facility. Similar ettringite trends were observed in the interior segments of the core samples. The compressive strength of the concrete at the mid-wall of the basin was measured, and similar microstructural analysis was conducted on these materials post compression testing. The microstructure was determined to be similar to near-surface segments of the core samples. The average strength was 4148 psi, which is well-above the design strength of 2500 psi. The analyses showed that phase alterations and minor cracking in a microstructure did not affect the design specification for the concrete

Closure Report for Corrective Action Unit 254: Area 25, R-MAD Decontamination Facility, Nevada Test Site, Nevada

Energy Technology Data Exchange (ETDEWEB)

G. N. Doyle

2002-02-01

Corrective Action Unit (CAU) 254 is located in Area 25 of the Nevada Test Site (NTS), approximately 100 kilometers (km) (62 miles) northwest of Las Vegas, Nevada. The site is located within the Reactor Maintenance, Assembly and Disassembly (R-MAD) compound and consists of Building 3126, two outdoor decontamination pads, and surrounding areas within an existing fenced area measuring approximately 50 x 37 meters (160 x 120 feet). The site was used from the early 1960s to the early 1970s as part of the Nuclear Rocket Development Station program to decontaminate test-car hardware and tooling. The site was reactivated in the early 1980s to decontaminate a radiologically contaminated military tank. This Closure Report (CR) describes the closure activities performed to allow un-restricted release of the R-MAD Decontamination Facility.

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

International Nuclear Information System (INIS)

ROBERT, HIERGESELL

2005-01-01

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

Radiation damage evaluation on concrete within a facility for Selective Production of Exotic Species (SPES Project), Italy

International Nuclear Information System (INIS)

Pomaro, B.; Salomoni, V.A.; Gramegna, F.; Prete, G.; Majorana, C.E.

2011-01-01

Highlights: → We present the effect of radiation on concrete as shielding material. → The coupling between hydro-thermal-mechanical fields and radiation damage is shown. → Attention is focused on numerical modelling of concrete in 3D domains. → A new estimate of the radiation damage parameter is given. → A risk assessment of concrete-radiation interactions is developed. - Abstract: Concrete is commonly used as a biological shield against nuclear radiation. As long as, in the design of nuclear facilities, its load carrying capacity is required together with its shielding properties, changes in the mechanical properties due to nuclear radiation are of particular significance and may have to be taken into account in such circumstances. The study presented here allows for reaching first evidences on the behavior of concrete when exposed to nuclear radiation in order to evaluate the consequent effect on the mechanical field, by means of a proper definition of the radiation damage, strictly connected with the strength properties of the building material. Experimental evidences on the decay of the mechanical modulus of concrete have allowed for implementing the required damage law within a 3D F.E. research code which accounts for the coupling among moisture, heat transfer and the mechanical field in concrete treated as a fully coupled porous medium. The development of the damage front in a concrete shielding wall is analyzed under neutron radiation and results within the wall thickness are reported for long-term radiation spans and several concrete mixtures in order to discuss the resulting shielding properties.

Radiation damage evaluation on concrete within a facility for Selective Production of Exotic Species (SPES Project), Italy.

Science.gov (United States)

Pomaro, B; Salomoni, V A; Gramegna, F; Prete, G; Majorana, C E

2011-10-30

Concrete is commonly used as a biological shield against nuclear radiation. As long as, in the design of nuclear facilities, its load carrying capacity is required together with its shielding properties, changes in the mechanical properties due to nuclear radiation are of particular significance and may have to be taken into account in such circumstances. The study presented here allows for reaching first evidences on the behavior of concrete when exposed to nuclear radiation in order to evaluate the consequent effect on the mechanical field, by means of a proper definition of the radiation damage, strictly connected with the strength properties of the building material. Experimental evidences on the decay of the mechanical modulus of concrete have allowed for implementing the required damage law within a 3D F.E. research code which accounts for the coupling among moisture, heat transfer and the mechanical field in concrete treated as a fully coupled porous medium. The development of the damage front in a concrete shielding wall is analyzed under neutron radiation and results within the wall thickness are reported for long-term radiation spans and several concrete mixtures in order to discuss the resulting shielding properties. Copyright © 2011 Elsevier B.V. All rights reserved.

The Barrier code for predicting long-term concrete performance

International Nuclear Information System (INIS)

Shuman, R.; Rogers, V.C.; Shaw, R.A.

1989-01-01

There are numerous features incorporated into a LLW disposal facility that deal directly with critical safety objectives required by the NRC in 10 CFR 61. Engineered barriers or structures incorporating concrete are commonly being considered for waste disposal facilities. The Barrier computer code calculates the long-term degradation of concrete structures in LLW disposal facilities. It couples this degradation with water infiltration into the facility, nuclide leaching from the waste, contaminated water release from the facility, and associated doses to members of the critical population group. The concrete degradation methodology of Barrier is described

Closure Plan for Active Low Level Burial Grounds

Energy Technology Data Exchange (ETDEWEB)

SKELLY, W.A.

2000-11-16

This plan has been prepared in response to direction from the U.S. Department of Energy. The purpose of the plan is to define approaches that will be implemented to ensure protection of the public and the environment when active Low-Level Burial Grounds (LLBGs) at the Hanford Site are closed. Performance assessments for active burial grounds in the 200 East and West 200 Areas provide current estimates of potential environmental contamination and doses to the ''maximum exposed individual'' from burial ground operation and closure and compare dose estimates to performance objective dose limits for the facilities. This is an Operational Closure Plan. The intent of the guidance in DOE Order 435.1 is that this plan will be a living document, like the facility performance assessments, and will be revised periodically through the operational life of the LLBGs to reflect updated information on waste inventory. management practices, facility transition planning, schedule dates, assessments of post-closure performance, and environmental consequences. Out year dates identified in this plan are tentative. A Final Closure Plan will be prepared in the future when the timing and extent of closure-related activities for LLBGs can be established with greater certainty. After current operations at the LLBGs are concluded, this plan proposes transitioning of these facilities to the Environmental Restoration Program. This action will enable the Environmental Restoration Program to design and implement consistent and coordinated final remedial actions for active and inactive LLBGs. Active and inactive burial grounds in the 200 West and 200 East Areas are commingled. This plan describes approaches that will be implemented during Interim Closure, Final Closure, and Institutional Control Periods to prepare LLBGs for surface barriers, and the construction of barriers, as well as the scope of inspection, monitoring and maintenance practices that will be performed during

Design, production and initial state of the closure

International Nuclear Information System (INIS)

2010-12-01

The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The production reports are included in the safety report for the KBS-3 repository and repository facility. The report provides input on the initial state of the closure and plugs in underground openings other than deposition tunnels for the assessment of the long-term safety, SR-Site. The initial state refers to the properties of the engineered barriers once they have been finally placed in the KBS-3 repository and will not be further handled within the repository facility. In addition, the report provides some input to the operational safety report, SR-Operation, on how the closure and plugs shall be handled and installed. The report presents the design premises and reference designs of the closure and plugs and verifies their conformity to the design premises. It also briefly deals with the production of the closure and plugs. Finally, the initial state of the closure and plugs and their conformity to the reference designs and design premises are presented

Design, production and initial state of the closure

Energy Technology Data Exchange (ETDEWEB)

2010-12-15

The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The production reports are included in the safety report for the KBS-3 repository and repository facility. The report provides input on the initial state of the closure and plugs in underground openings other than deposition tunnels for the assessment of the long-term safety, SR-Site. The initial state refers to the properties of the engineered barriers once they have been finally placed in the KBS-3 repository and will not be further handled within the repository facility. In addition, the report provides some input to the operational safety report, SR-Operation, on how the closure and plugs shall be handled and installed. The report presents the design premises and reference designs of the closure and plugs and verifies their conformity to the design premises. It also briefly deals with the production of the closure and plugs. Finally, the initial state of the closure and plugs and their conformity to the reference designs and design premises are presented

300 Area waste acid treatment system closure plan

International Nuclear Information System (INIS)

LUKE, S.N.

1999-01-01

The Hanford Facility Dangerous Waste Permit Application is considered to be a single application organized into a General Information Portion (document number DOERL-91-28) and a Unit-Specific Portion. The scope of the Unit-Specific Portion includes closure plan documentation submitted for individual, treatment, storage, and/or disposal units undergoing closure, such as the 300 Area Waste Acid Treatment System. Documentation contained in the General Information Portion is broader in nature and could be used by multiple treatment, storage, and/or disposal units (e.g., the glossary provided in the General Information Portion). Whenever appropriate, 300 Area Waste Acid Treatment System documentation makes cross-reference to the General Information Portion, rather than duplicating text. This 300 Area Waste Acid Treatment System Closure Plan (Revision 2) includes a Hanford Facility Dangerous Waste Permit Application, Part A, Form 3. Information provided in this closure plan is current as of April 1999

300 Area waste acid treatment system closure plan

Energy Technology Data Exchange (ETDEWEB)

LUKE, S.N.

1999-05-17

The Hanford Facility Dangerous Waste Permit Application is considered to be a single application organized into a General Information Portion (document number DOERL-91-28) and a Unit-Specific Portion. The scope of the Unit-Specific Portion includes closure plan documentation submitted for individual, treatment, storage, and/or disposal units undergoing closure, such as the 300 Area Waste Acid Treatment System. Documentation contained in the General Information Portion is broader in nature and could be used by multiple treatment, storage, and/or disposal units (e.g., the glossary provided in the General Information Portion). Whenever appropriate, 300 Area Waste Acid Treatment System documentation makes cross-reference to the General Information Portion, rather than duplicating text. This 300 Area Waste Acid Treatment System Closure Plan (Revision 2) includes a Hanford Facility Dangerous Waste Permit Application, Part A, Form 3. Information provided in this closure plan is current as of April 1999.

Development of contaminated concrete removing system 'Clean cut method'

International Nuclear Information System (INIS)

Kinoshita, Takehiko; Tanaka, Tsutomu; Funakawa, Naoyoshi; Idemura, Hajime; Sakashita, Fumio; Tajitsu, Yoshiteru

1989-01-01

In the case of decommissioning nuclear facilities such as nuclear power stations, nuclear fuel facilities and RI handling facilities and carrying out reconstruction works, if there is radioactive contamination on the surfaces of concrete structures such as the floors and walls of the buildings for nuclear facilities, it must be removed. Since concrete is porous, contamination infiltrates into the inside of concrete, and the wiping of surfaces only or chemical decontamination cannot remove it, therefore in most cases, contaminated concrete must be removed. The removal of concrete surfaces has been carried out with chipping hammers, grinders and so on, but many problems arise due to it. In order to solve these problems, the mechanical cutting method was newly devised, and clean cut method (CCRS) was completed. The depth of cutting from concrete surface is set beforehand, and the part to be removed is accurately cut, at the same time, the concrete powder generated is collected nearly perfectly, and recovered into a drum. The outline of the method and the constitution of the system, the features of the clean cut method, the development of the technology for cutting concrete and the technology for recovering concrete powder, and the test of verifying decontamination are reported. (K.I.)

Closure Report for Corrective Action Unit 562: Waste Systems, Nevada National Security Site, Nevada

Energy Technology Data Exchange (ETDEWEB)

NSTec Environmental Restoration

2012-08-15

This Closure Report (CR) presents information supporting closure of Corrective Action Unit (CAU) 562, Waste Systems, and provides documentation supporting the completed corrective actions and confirmation that closure objectives for CAU 562 were met. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; the U.S. Department of Energy (DOE), Environmental Management; the U.S. Department of Defense; and DOE, Legacy Management (FFACO, 1996 as amended). CAU 562 consists of the following 13 Corrective Action Sites (CASs), located in Areas 2, 23, and 25 of the Nevada National Security Site: · CAS 02-26-11, Lead Shot · CAS 02-44-02, Paint Spills and French Drain · CAS 02-59-01, Septic System · CAS 02-60-01, Concrete Drain · CAS 02-60-02, French Drain · CAS 02-60-03, Steam Cleaning Drain · CAS 02-60-04, French Drain · CAS 02-60-05, French Drain · CAS 02-60-06, French Drain · CAS 02-60-07, French Drain · CAS 23-60-01, Mud Trap Drain and Outfall · CAS 23-99-06, Grease Trap · CAS 25-60-04, Building 3123 Outfalls Closure activities began in October 2011 and were completed in April 2012. Activities were conducted according to the Corrective Action Plan for CAU 562 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2011). The corrective actions included No Further Action and Clean Closure. Closure activities generated sanitary waste and hazardous waste. Some wastes exceeded land disposal limits and required offsite treatment prior to disposal. Other wastes met land disposal restrictions and were disposed in appropriate onsite or offsite landfills. NNSA/NSO requests the following: · A Notice of Completion from the Nevada Division of Environmental Protection to NNSA/NSO for closure of CAU 562 · The transfer of CAU 562 from Appendix III to Appendix IV, Closed Corrective Action Units, of the FFACO

Elevated temperature effects on concrete properties

International Nuclear Information System (INIS)

Grant, P.R.; Gruber, R.S.; Van Katwijk, C.

1993-08-01

The design of facilities to process or store radioactive wastes presents many challenging engineering problems. Such facilities must not only provide for safe storage of radioactive wastes but they must also be able to maintain confinement of these materials during and after natural phenomena events. Heat generated by the radioactive decay of the wastes will cause the temperature of the concrete containment structure to increase to a magnitude higher than that found in conventional structures. These elevated temperatures will cause strength-related concrete properties to degrade over time. For concrete temperatures less than 150 degree F, no reduction in strength is taken and the provisions of ACI 349, which states that higher temperatures are allowed if tests are provided to evaluate the reduction in concrete strength properties, apply. Methods proposed in a Pacific Northwest Laboratory (PNL) report, Modeling of Time-Variant Concrete Properties at Elevated Temperatures, can be used to evaluate the effects of elevated temperatures on concrete properties. Using these modified concrete properties the capacity of a concrete structure, subjected to elevated temperatures, to resist natural phenomena hazards can be determined

Infiltration experiment for closure cap evaluation at the Savannah River Site

International Nuclear Information System (INIS)

Roddy, N.S.; Cook, J.R.

1990-01-01

This document discusses several large waste disposal facilities at the Savannah River Site which are being closed. These facilities include two seepage basins and the low-level waste disposal facility. The key element of the closures is the construction of a cap system to limit the infiltration of water which might reach the disposed waste. Cap designs have been modeled using the Hydrologic Evaluation of Landfill Performance (HELP) computer code. This code was developed by the US Army Corps of Engineers for the Environmental Protection Agency to model the effects of various cap and liner designs on the water balance at landfills. A field experiment has been set up which will allow the results of the HELP Code to be verified at the Savannah River Site (SRS) by measuring the actual water balance created by closure cap configurations which will be used in waste site closures at SRS. Two of the caps will be similar to those used for the planned closure activities. Each one has a specific closure arrangement. Once operational, the experiment will be evaluated for a five-year period

Repository Closure and Sealing Approach

International Nuclear Information System (INIS)

A.T. Watkins

2000-01-01

The scope of this analysis will be to develop the conceptual design of the closure seals and their locations in the Subsurface Facilities. The design will be based on the recently established program requirements for transitioning to the Site Recommendation (SR) design as outlined by ''Approach to Implementing the Site Recommendation Baseline'' (Stroupe 2000) and the ''Monitored Geologic Repository Project Description Document'' (CRWMS M andO 1999b). The objective of this analysis will be to assist in providing a description for the Subsurface Facilities System Description Document, Section 2 and finally to document any conclusions reached in order to contribute and provide support to the SR. This analysis is at a conceptual level and is considered adequate to support the SR design. The final closure barriers and seals for the ventilation shafts, and the north and south ramps will require these openings to be permanently sealed to limit excessive air and water inflows and prevent human intrusion. The major tasks identified with closure in this analysis are: (1) Developing the overall subsurface seal layout and identifying design and operational interfaces for the Subsurface Facilities. (2) Summarizing the general site conditions and general rock characteristic with respect to seal location and describing the seal selected. (3) Identify seal construction materials, methodology of construction and strategic locations including design of the seal and plugs. (4) Discussing methods to prevent human intrusion

Residual radioactivity investigation and radiological assessments for self-disposal of concrete waste in nuclear fuel processing facility

International Nuclear Information System (INIS)

Seol, Jeung Gun; Ryu, Jae Bong; Cho, Suk Ju; Yoo, Sung Hyun; Song, Jung Ho; Baek, Hoon; Kim, Seong Hwan; Shin, Jin Seong; Park, Hyun Kyoun

2007-01-01

In this study, domestic regulatory requirement was investigated for self-disposal of concrete waste from nuclear fuel processing facility. And after self-disposal as landfill or recycling/reuse, the exposure dose was evaluated by RESRAD Ver. 6.3 and RESRAD BUILD Ver. 3.3 computing code for radiological assessments of the general public. Derived clearance level by the result of assessments for the exposure dose of the general public is 0.1071Bq/g (3.5% enriched uranium) for landfill and 0.05515 Bq/cm 2 (5% enriched uranium) for recycling/reuse respectively. Also, residual radioactivity of concrete waste after decontamination was investigated in this study. The result of surface activity is 0.01Bq/cm 2 for emitter and the result of radionuclide analysis for taken concrete samples from surface of concrete waste is 0.0297Bq/g for concentration of 238 U, below 2w/o for enrichment of 235 U and 0.0089Bq/g for artificial contamination of 238 U respectively. Therefore, radiological hazard of concrete waste by self-disposal as landfill and recycling/reuse is below clearance level to comply with clearance criterion provided for Notice No. 2001-30 of the MOST and Korea Atomic Energy Act

Evaluation of precast concrete slabs using a heavy vehicle simulator

CSIR Research Space (South Africa)

Kohler, E

2008-10-01

Full Text Available Precast slabs are considered an attractive pavement option for rehabilitation or reconstruction cases where traffic closures of less than eight hours are required. Benefits include long life expectancy of concrete cast in factory...

Special protective concretes

International Nuclear Information System (INIS)

Bouniol, P.

2001-01-01

Concrete is the most convenient material when large-scale radiation protection is needed. Thus, special concretes for nuclear purposes are used in various facilities like reactors, reprocessing centers, storage sites, accelerators, hospitals with nuclear medicine equipment, food ionization centers etc.. The recent advances made in civil engineering for the improvement of concrete durability and compactness are for a large part transposable to protection concretes. This article presents the basic knowledge about protection concretes with the associated typological and technological aspects. A large part is devoted to the intrinsic properties of concretes and to their behaviour in irradiation and temperature conditions: 1 - definition and field of application of special protective concretes; 2 - evolution of concepts and technologies (durability of structures, techniques of formulation, new additives, market evolution); 3 - design of protective structures (preliminary study, radiation characteristics, thermal constraints, damping and dimensioning, mechanical criteria); 4 - formulation of special concretes (general principles, granulates, hydraulic binders, pulverulent additives, water/cement ratio, reference composition of some special concretes); 5 - properties of special concretes (damping and thermo-mechanical properties); 6 - induced-irradiation and temperature phenomena (activation, radiolysis, mineralogical transformations, drying, shrinking, creep, corrosion of reinforcement). (J.S.)

Evaluation of the effects of strain rate on material properties of the high strength concrete used in nuclear facilities

International Nuclear Information System (INIS)

Kawaguchi, Shohei; Shirai, Koji; Takayanagi, Hideaki

2011-01-01

Concrete physical properties (compressive strength, tensile strength, initial elastic modulus and maximum strain) affected by strain rate weren't fully utilize for material model in dynamic response analysis for seismic and impact load because of few reports and various difficulties of impact tests. Split Hopkinson Pressure Bar (SHPB) methods are the most popular high-speed material testing and were also applied for composite material. We applied SHPB for concrete specimen and reported the strain rate effect to the concrete physical property. We used hydraulic testing device for 10 -5 /s to 10 0 /s strain rate and SHPB methods for over 10 1 /s. Four cases of concrete tests (high (50MPa at 28days)/low (35MPa at 28days) compressive strength (based on the test of exiting nuclear power facilities) and dry/wet condition) were done. And we formulated strain rate effect about compressive strength and initial elastic modulus from comparing with previous studies. (author)

The Environmental Agency's Assessment of the Post-Closure Safety Case for the BNFL DRIGG Low Level Radioactive Waste Disposal Facility

International Nuclear Information System (INIS)

Streatfield, I. J.; Duerden, S. L.; Yearsley, R. A.

2002-01-01

The Environment Agency is responsible, in England and Wales, for authorization of radioactive waste disposal under the Radioactive Substances Act 1993. British Nuclear Fuels plc (BNFL) is currently authorized by the Environment Agency to dispose of solid low level radioactive waste at its site at Drigg, near Sellafield, NW England. As part of a planned review of this authorization, the Environment Agency is currently undertaking an assessment of BNFL's Post-Closure Safety Case Development Programme for the Drigg disposal facility. This paper presents an outline of the review methodology developed and implemented by the Environment Agency specifically for the planned review of BNFL's Post-Closure Safety Case. The paper also provides an overview of the Environment Agency's progress in its on-going assessment programme

Addendum to the Closure Report for Corrective Action Unit 113: Area 25 R-MAD Facility, Nevada National Security Site, Nevada

International Nuclear Information System (INIS)

2011-01-01

This addendum to the Closure Report for Corrective Action Unit 113: Area 25, Reactor Maintenance, Assembly, and Disassembly Facility, Building 3110, Nevada Test Site, Nevada, DOE/NV--891-VOL I-Rev. 1, dated July 2003, provides details of demolition, waste disposal, and use restriction (UR) modification for Corrective Action Unit 113, Area 25 R-MAD Facility. Demolition was completed on July 15, 2010, when the last of the building debris was disposed. Final field activities were concluded on August 30, 2010, after all equipment was demobilized and UR signs were posted. This work was funded by the American Recovery and Reinvestment Act.

Evolution of sorption properties in large-scale concrete structures accounting for long-term physical-chemical concrete degradation - 59297

International Nuclear Information System (INIS)

Perko, Janez; Jacques, Diederik; Mallants, Dirk; Seetharam, Suresh

2012-01-01

Long-term safety of radioactive waste disposal facilities relies on the longevity of natural or engineered barriers designed to minimize the migration of contaminants from the facility into the environment. Especially near surface disposal facilities, such as planned by ONDRAF/NIRAS for the Dessel site in Belgium, long-term safety relies almost exclusively on the containment ability of the engineered barriers (EB) with concrete being the most important EB material used. Concrete is preferred over other materials mainly due to its favourable chemical properties resulting in a high chemical retention capacity, and owing to its good hydraulic isolation properties. However, due to the long time frames typically involved in safety assessment, the chemical, physical and mechanical properties of concrete evolve in time. The alterations in concrete mineralogy also cause changes in pH and sorption behaviour for many radionuclides during chemical degradation processes. Application of dynamic sorption of concrete requires an adequate knowledge of long-term concrete degradation processes, knowledge of the effect of changing mineralogy to sorption of radionuclides and knowledge of large-scale system behaviour over time. Moreover, when applied to safety assessment models, special attention is required to assure robustness and transparency of the implementation. The discussion in this paper focuses on the sorption properties of concrete; selection of data, rescaling issues and on the hypotheses used to build a robust and yet transparent dynamic model for large-scale concrete structures for assessing the long-term performance. In this paper we summarize the steps required for the appropriate use of sorption values for large-scale cementitious components accounting for long-term concrete degradation in safety assessment studies. Four steps were recognized through the safety assessment in the framework of the license application for the near-surface disposal facility in Dessel

Closure of a mixed waste landfill: Lessons learned

International Nuclear Information System (INIS)

Phifer, M.A.

1990-01-01

Much experience has been gained during the closure of the Mixed Waste Management Facility (MWMF) at the Savannah River Site (SRS) and many lessons were learned. This knowledge was applied to other closures at SRS yielding decreased costs, schedule enhancement, and increased overall project efficiency. The next major area of experience to be gained at SRS in the field of waste site closures will be in the upkeep, maintenance, and monitoring of clay caps. Further test programs will be required to address these requirements

Light Water Reactor Sustainability Program: survey of models for concrete degradation

International Nuclear Information System (INIS)

2014-01-01

Concrete has been used in the construction of nuclear facilities because of two primary properties: its structural strength and its ability to shield radiation. Concrete structures have been known to last for hundreds of years, but they are also known to deteriorate in very short periods of time under adverse conditions. The use of concrete in nuclear facilities for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility. The goal of this report is to review and document the main aging mechanisms of concern for concrete structures in nuclear power plants (NPPs) and the models used in simulations of concrete aging and structural response of degraded concrete structures. This is in preparation for future work to develop and apply models for aging processes and response of aged NPP concrete structures in the Grizzly code. To that end, this report also provides recommendations for developing more robust predictive models for aging effects of performance of concrete.

Reliability assessment of underground shaft closure

International Nuclear Information System (INIS)

Fossum, A.F.; Munson, D.E.

1994-01-01

The intent of the WIPP, being constructed in the bedded geologic salt deposits of Southeastern New Mexico, is to provide the technological basis for the safe disposal of radioactive Transuranic (TRU) wastes generated by the defense programs of the United States. In determining this technological basis, advanced reliability and structural analysis techniques are used to determine the probability of time-to-closure of a hypothetical underground shaft located in an argillaceous salt formation and filled with compacted crushed salt. Before being filled with crushed salt for sealing, the shaft provides access to an underground facility. Reliable closure of the shaft depends upon the sealing of the shaft through creep closure and recompaction of crushed backfill. Appropriate methods are demonstrated to calculate cumulative distribution functions of the closure based on laboratory determined random variable uncertainty in salt creep properties

Concrete spaller. Innovative technology summary report

International Nuclear Information System (INIS)

1998-12-01

The US Department of Energy (DOE) has numerous buildings and facilities that have become contaminated through operation of nuclear reactors, fuel fabrication processes, and research laboratory operations. These buildings and facilities, often constructed of concrete, need to be decontaminated before they can be safely decommissioned or demolished. Pacific Northwest National Laboratory's concrete spaller is a hand-held tool that can be used for decontaminating flat or slightly curved concrete surfaces, obtaining concrete samples, and in-depth removal from cracks in concrete. The concrete spaller includes a 9-ton hydraulic cylinder and spalling bit. It runs from a hydraulic pump that expands the spaller in pre-drilled holes in the concrete. The result is removal of concrete chunks that fall into the attached metal shroud. The concrete spaller is more efficient than traditional tools such as hand-held pneumatic scabblers and scalers. For example, the spaller is capable of spalling 1.3 m 3 /hr (0.23 ft 2 /min), compared to 1.1 m 2 /hr (0.20 ft 2 /min), for the baseline scabbler and scaler demonstrated at 3-mm (1/8-in.) depth. The spaller is also capable of removing concrete at a greater depth than traditional tools. Operating cost of the spaller ($128/m 2 or $11.93/ft 2 [optimum conditions]) is less than the baseline tools: scaler ($155/m 2 or $14.40/ft 2 ) and scabbler ($156/m 2 or $14.53/ft 2 )

Incorporating Sustainability into Site Closure - A Field Example

Science.gov (United States)

Austrins, L. M.; West, J.

2013-12-01

Long term management of former chemical production facilities can be a costly and time consuming element of site closure, however, implementation of creative measures to introduce sustainability and reduce the need for onsite presence can be successfully incorporated into the site closure process. A case study demonstrating this involves a facility located in Sarnia, Ontario, which was an active multi chemical production facility from the 1940s, until it was decommissioned and sold between 2005 and 2010. The facility consisted of 322 acres of production areas. Several elements which allowed for reduced onsite presence and lower management costs were incorporated into the site decommissioning plan, including; phased remediation planning, and selection of sustainable components as part of remediation, surface water management, and groundwater management. The sustainability and management modifications were successfully negotiated and approved by the local regulatory agency. Due to the size and complexity of the site, a holistic approach for the facility was needed and included the development of a comprehensive decision matrix. Each remediation alternative incorporated sustainable practices. Ex-situ remediation consisted of excavation of contaminated subsurface medium and consolidation at a 4.7 acre onsite soil treatment area designed specifically for the site closure process. In-situ remediation consisted of injection of amendment into the native soils using hydraulic fracture and injection. When the plant was an active operating facility, groundwater management required active pumping and groundwater treatment through a series of carbon treatment units. Active pumping has been replaced by passive hydraulic control through the use of tree plantations.

Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

International Nuclear Information System (INIS)

Michael R, Kruzic

2008-01-01

Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility (Figure 1) was used in the early to mid-1960s for testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles. The TCA facility, known as Corrective Action Unit 115, was decontaminated and decommissioned (D and D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the Federal Facility Agreement and Consent Order. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously, provided technical decisions are made by an experienced decision maker within the site conceptual site model. Facility closure involved a seven-step decommissioning strategy. First, preliminary investigation activities were performed, including review of process knowledge documentation, targeted facility radiological and hazardous material surveys, concrete core drilling and analysis, shield wall radiological characterization, and discrete sampling, which proved to be very useful and cost-effective in subsequent decommissioning planning and execution and worker safety. Second, site setup and mobilization of equipment and personnel were completed. Third, early removal of hazardous materials, including asbestos, lead, cadmium, and oil, was performed ensuring worker safety during more invasive demolition activities. Process piping was to be verified void of contents. Electrical systems were de-energized and other systems were rendered free of residual energy. Fourth, areas of high radiological contamination were decontaminated using multiple methods. Contamination levels varied across the facility. Fixed beta/gamma contamination levels ranged up to 2 million disintegrations per minute (dpm)/100

Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

Energy Technology Data Exchange (ETDEWEB)

Michael R. Kruzic

2008-06-01

Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility (Figure 1) was used in the early to mid-1960s for testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles. The TCA facility, known as Corrective Action Unit 115, was decontaminated and decommissioned (D&D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the Federal Facility Agreement and Consent Order. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously, provided technical decisions are made by an experienced decision maker within the site conceptual site model. Facility closure involved a seven-step decommissioning strategy. First, preliminary investigation activities were performed, including review of process knowledge documentation, targeted facility radiological and hazardous material surveys, concrete core drilling and analysis, shield wall radiological characterization, and discrete sampling, which proved to be very useful and cost-effective in subsequent decommissioning planning and execution and worker safety. Second, site setup and mobilization of equipment and personnel were completed. Third, early removal of hazardous materials, including asbestos, lead, cadmium, and oil, was performed ensuring worker safety during more invasive demolition activities. Process piping was to be verified void of contents. Electrical systems were de-energized and other systems were rendered free of residual energy. Fourth, areas of high radiological contamination were decontaminated using multiple methods. Contamination levels varied across the facility. Fixed beta/gamma contamination levels ranged up to 2 million disintegrations per minute (dpm)/100

Closure Report for the 92-Acre Area and Corrective Action Unit 111: Area 5 WMD Retired Mixed Waste Pits, Nevada National Security Site, Nevada

Energy Technology Data Exchange (ETDEWEB)

NSTec Environmental Restoration

2012-02-21

This Closure Report (CR) presents information supporting closure of the 92-Acre Area, which includes Corrective Action Unit (CAU) 111, 'Area 5 WMD Retired Mixed Waste Pits.' This CR provides documentation supporting the completed corrective actions and confirmation that the closure objectives were met. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) (FFACO, 1996 [as amended March 2010]). Closure activities began in January 2011 and were completed in January 2012. Closure activities were conducted according to Revision 1 of the Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) for the 92-Acre Area and CAU 111 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2010). The following closure activities were performed: (1) Construct an engineered evapotranspiration cover over the boreholes, trenches, and pits in the 92-Acre Area; (2) Install use restriction (UR) warning signs, concrete monuments, and subsidence survey monuments; and (3) Establish vegetation on the covers. UR documentation is included as Appendix C of this report. The post-closure plan is presented in detail in Revision 1 of the CADD/CAP for the 92-Acre Area and CAU 111, and the requirements are summarized in Section 5.2 of this document. When the next request for modification of Resource Conservation and Recovery Act Permit NEV HW0101 is submitted to the Nevada Division of Environmental Protection (NDEP), the requirements for post-closure monitoring of the 92-Acre Area will be included. NNSA/NSO requests the following: (1) A Notice of Completion from NDEP to NNSA/NSO for closure of CAU 111; and (2) The transfer of CAU 111 from Appendix III to Appendix IV, Closed Corrective Action Units, of the FFACO.

Closure Report for the 92-Acre Area and Corrective Action Unit 111: Area 5 WMD Retired Mixed Waste Pits, Nevada National Security Site, Nevada

International Nuclear Information System (INIS)

2012-01-01

This Closure Report (CR) presents information supporting closure of the 92-Acre Area, which includes Corrective Action Unit (CAU) 111, 'Area 5 WMD Retired Mixed Waste Pits.' This CR provides documentation supporting the completed corrective actions and confirmation that the closure objectives were met. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) (FFACO, 1996 (as amended March 2010)). Closure activities began in January 2011 and were completed in January 2012. Closure activities were conducted according to Revision 1 of the Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) for the 92-Acre Area and CAU 111 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO), 2010). The following closure activities were performed: (1) Construct an engineered evapotranspiration cover over the boreholes, trenches, and pits in the 92-Acre Area; (2) Install use restriction (UR) warning signs, concrete monuments, and subsidence survey monuments; and (3) Establish vegetation on the covers. UR documentation is included as Appendix C of this report. The post-closure plan is presented in detail in Revision 1 of the CADD/CAP for the 92-Acre Area and CAU 111, and the requirements are summarized in Section 5.2 of this document. When the next request for modification of Resource Conservation and Recovery Act Permit NEV HW0101 is submitted to the Nevada Division of Environmental Protection (NDEP), the requirements for post-closure monitoring of the 92-Acre Area will be included. NNSA/NSO requests the following: (1) A Notice of Completion from NDEP to NNSA/NSO for closure of CAU 111; and (2) The transfer of CAU 111 from Appendix III to Appendix IV, Closed Corrective Action Units, of the FFACO.

POST-CLOSURE INSPECTION REPORT FOR CORRECTIVE ACTION UNIT 92: AREA 6 DECON PAD FACILITY, NEVADA. TEST SITE NEVADA, FOR THE PERIOD JANUARY 2004 - DECEMBER 2004

International Nuclear Information System (INIS)

BECHTEL NEVADA

2005-01-01

This Post-Closure Inspection Report provides an analysis and summary of inspections for Corrective Action Unit (CAU) 92, Area 6 Decon Pond Facility, Nevada Test Site, Nevada. CAU 92 was closed in accordance with the Resource Conservation and Recovery Act (RCRA) Part B Operational Permit (Nevada Division of Environmental Protection, 1995) and the Federal Facility Agreement and Consent Order of 1996 on May 11, 1999. CAU 92 consists of two Corrective Action Sites (CASs): CAS 06-04-01, Decon Pad oil/Water Separator; and CAS 06-05-02, Decontamination Pond (RCRA). Both CASs have use restrictions; however, only CAS 06-05-02, Decontamination Pond (RCRA), requires post-closure inspections. CAS 06-04-01, Decon Pad Oil/Water Separator, is located inside the fence at the Building 6-605 compound. This report covers the annual period January 2004 through December 2004

POST-CLOSURE INSPECTION REPORT FOR CORRECTIVE ACTION UNIT 92: AREA 6 DECON POND FACILITY, NEVADA TEST SITE, NEVADA FOR CALENDAR YEAR 2005

International Nuclear Information System (INIS)

2006-01-01

This Post-Closure Inspection Report provides an analysis and summary of inspections for Corrective Action Unit (CAU) 92, Area 6 Decon Pond Facility, Nevada Test Site, Nevada. CAU 92 was closed in accordance with the Resource Conservation and Recovery Act (RCRA) Part B Operational Permit (Nevada Division of Environmental Protection (NDEP), 1995) and the Federal Facility Agreement and Consent Order of 1996. Closure activities were completed on February 16, 1999, and the Closure Report (U.S. Department of Energy, Nevada Operations Office, 1999) was approved and a Notice of Completion issued by the NDEP on May 11, 1999. CAU 92 consists of two Corrective Action Sites (CASs): CAS 06-04-01, Decon Pad Oil/Water Separator; and CAS 06-05-02, Decontamination Pond (RCRA). Both CASs have use restrictions; however, only CAS 06-05-02 requires post-closure inspections. Visual inspections of the cover and fencing at CAS 06-05-02 are performed quarterly. Additional inspections are conducted if precipitation occurs in excess of 1.28 centimeters (cm) (0.50 inches [in]) in a 24-hour period. This report covers calendar year 2005. Quarterly site inspections were performed in March, June, September, and December of 2005. All observations indicated the continued integrity of the unit. No issues or concerns were noted, and no corrective actions were necessary. Copies of the inspection checklists and field notes completed during each inspection are included in Appendix A. Five additional inspections were performed after precipitation events that exceeded 1.28 cm (0.50 in) within a 24-hour period during 2005. No significant changes in site conditions were noted during these inspections, and no corrective actions were necessary. Copies of the inspection checklists and field notes completed during each inspection are included in Appendix A. Precipitation records for 2005 are included in Appendix C

Interim Status Closure Plan Open Burning Treatment Unit Technical Area 16-399 Burn Tray

Energy Technology Data Exchange (ETDEWEB)

Vigil-Holterman, Luciana R. [Los Alamos National Laboratory

2012-05-07

This closure plan describes the activities necessary to close one of the interim status hazardous waste open burning treatment units at Technical Area (TA) 16 at the Los Alamos National Laboratory (LANL or the Facility), hereinafter referred to as the 'TA-16-399 Burn Tray' or 'the unit'. The information provided in this closure plan addresses the closure requirements specified in the Code of Federal Regulations (CFR), Title 40, Part 265, Subparts G and P for the thermal treatment units operated at the Facility under the Resource Conservation and Recovery Act (RCRA) and the New Mexico Hazardous Waste Act. Closure of the open burning treatment unit will be completed in accordance with Section 4.1 of this closure plan.

Enhanced sealing project (ESP): evolution of a full-sized concrete and bentonite shaft seal

International Nuclear Information System (INIS)

Dixon, D.A.; Priyanto, D.G.; Martino, J.B.; De Combarieu, M.; Johansson, R.; Korkeakoski, P.; Villagran, J.

2012-01-01

Document available in extended abstract form only. A full-scale shaft seal was designed and installed in the 5-m-diameter access shaft at Atomic Energy of Canada's (AECL's) Underground Research Laboratory (URL) at the point where he shaft intersects an ancient water-bearing, low-angle thrust fault in granitic rock. The seal, part of the permanent closure of the URL, consists of a 6-m-thick bentonite-based component sandwiched between 3-m-thick upper and lower, 3-m-thick concrete components. The bentonite-based component spans the fracture zone and extends approximately 1 m beyond the maximum identified extent of the fracture. This design was adopted in order to limit water from the deeper, saline regions mixing with the fresher, near-surface groundwater regime. The concrete components provide the mechanical confinement and an in situ compacted 40/60 mixture of bentonite clay and quartz sand provides the sealing component. Construction of the shaft seal was done as part of Canada's Nuclear Legacies Liability Program. However, monitoring the seal evolution was not part of the decommissioning program's mandate. In addition to accomplishing the permanent closure of the URL, this seal's construction provides a unique opportunity to instrument and monitor the evolution of a full-scale shaft seal as well as the recovery of the regional groundwater regime as the facility passively floods above the seal. A jointly funded monitoring project was developed by NWMO (Canada), SKB (Sweden), Posiva (Finland) and ANDRA (France) and since mid 2009 the thermal, hydraulic and mechanical evolution of the seal has been constantly monitored. The evolution of the type of seal being monitored in the ESP is of relevance to repository closure planning, gaining confidence in the functionality of shaft seals. Although constructed in a crystalline rock medium, the results of the ESP are expected to be relevant to the performance of seals in a variety of host rock types. The shaft seal has been

Investigation on integrity of JMTR concrete structures, cooling system and utility facilities

International Nuclear Information System (INIS)

Ebisawa, Hiroyuki; Tobita, Kenji; Fukasaku, Akitomi; Kaminaga, Masanori

2010-02-01

The condition of facilities and components to be used for re-operation of the Japan Materials Testing Reactor (JMTR) from FY2011, was investigated before the refurbishment work. An investigation of aged components (aged-investigation) was carried out for concrete structures of the JMTR reactor building, exhaust stack, trench, canal, filter banks and for aged components of tanks in the primary cooling system, heat exchangers, pipes in the secondary cooling system, cooling tower, emergency generators and so on, in order to identify their integrity. The aged-investigation was carried out from the beginning of FY2007. As a result, cracks of concrete structures such as the exhaust stack, a foundation of the UCL (Utility Cooling Line) elevated water tank were repaired and pipe linings of secondary cooling system were replaced. Motors of primary cooling pumps, pumps in the secondary cooling system and in other systems were decided to replace from viewpoints of future maintenance and improvement of reliability. Other components and the reactor building were decided to use continuously for a long-term by appropriate maintenance activities based on the long-term maintenance plan. In this paper, the aged-investigation for the JMTR reactor building, heat exchangers and emergency generators is presented. (author)

Closure Report for Corrective Action Unit 135: Areas 25 Underground Storage Tanks, Nevada Test Site, Nevada

Energy Technology Data Exchange (ETDEWEB)

D. H. Cox

2001-06-01

Corrective Action Unit (CAU) 135, Area 25 Underground Storage Tanks, was closed in accordance with the approved Corrective Action Plan (DOE/NV, 2000). CAU 135 consists of three Corrective Action Sites (CAS). Two of these CAS's were identified in the Corrective Action Investigation Data Quality Objective meeting as being improperly identified as underground storage tanks. CAS 25-02-03 identified as the Deluge Valve Pit was actually an underground electrical vault and CAS 25-02-10 identified as an Underground Storage Tank was actually a former above ground storage tank filled with demineralized water. Both of these CAS's are recommended for a no further action closure. CAS 25-02-01 the Underground Storage Tanks commonly referred to as the Engine Maintenance Assembly and Disassembly Waste Holdup Tanks and Vault was closed by decontaminating the vault structure and conducting a radiological verification survey to document compliance with the Nevada Test Site unrestricted use release criteria. The Area 25 Underground Storage Tanks, (CAS 25-02-01), referred to as the Engine Maintenance, Assembly, and Disassembly (E-MAD) Waste Holdup Tanks and Vault, were used to receive liquid waste from all of the radioactive and cell service area drains at the E-MAD Facility. Based on the results of the Corrective Action Investigation conducted in June 1999, discussed in ''The Corrective Action Investigation Plan for Corrective Action Unit 135: Area 25 Underground Storage Tanks, Nevada Test Site, Nevada'' (DOE/NV, 199a), one sample from the radiological survey of the concrete vault interior exceeded radionuclide preliminary action levels. The analytes from the sediment samples exceeded the preliminary action levels for polychlorinated biphenyls, Resource Conservation and Recovery Act metals, total petroleum hydrocarbons as diesel-range organics, and radionuclides. The CAU 135 closure activities consisted of scabbling radiological ''hot spots

The feasibility of recycling contaminated concrete

International Nuclear Information System (INIS)

Ayers, K.W.; Corroon, W.; Parker, F.L.

1999-01-01

The changing mission of the Department of Energy along with the aging of many of its facilities has resulted in renewed emphasis on decontaminating and decommissioning surplus structures. Currently DOE is decontaminating some concrete and sending the clean material to C and D disposal facilities. In other instance, DOE is sending contaminated concrete to LLW disposal facilities. This paper examines the economic feasibility of decontaminating the concrete and recycling the rubble as clean aggregate. A probabilistic cost model was used to examine six potential recycling and disposal scenarios. The model predicted potential costs saving across the DOE complex of nearly one billion dollars. The ability of local markets to assimilate the recycled material was estimated for Washington, Idaho, Tennessee, New Mexico, and South Carolina. The relationships between a number of the economic model's variables were examined to develop operating ranges for initial managerial evaluation of recycling

Light Water Reactor Sustainability Program: Survey of Models for Concrete Degradation

Energy Technology Data Exchange (ETDEWEB)

Spencer, Benjamin W. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Huang, Hai [Idaho National Lab. (INL), Idaho Falls, ID (United States). Energy and Environment Science and Technology

2014-08-01

Concrete is widely used in the construction of nuclear facilities because of its structural strength and its ability to shield radiation. The use of concrete in nuclear facilities for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility. As such, when life extension is considered for nuclear power plants, it is critical to have predictive tools to address concerns related to aging processes of concrete structures and the capacity of structures subjected to age-related degradation. The goal of this report is to review and document the main aging mechanisms of concern for concrete structures in nuclear power plants (NPPs) and the models used in simulations of concrete aging and structural response of degraded concrete structures. This is in preparation for future work to develop and apply models for aging processes and response of aged NPP concrete structures in the Grizzly code. To that end, this report also provides recommendations for developing more robust predictive models for aging effects of performance of concrete.

Longterm performance of structural component of intermediate- and low-level radioactive waste disposal facility

International Nuclear Information System (INIS)

Whang, J. H.; Kim, S. S.; Chun, T. H.; Lee, J. M.; Yum, M. O.; Kim, J. H.; Kim, M. S.

1997-03-01

Underground repository for intermediate- and low-level radioactive waste is to be sealed and closed after operation. Structural components, which are generally made of cement concrete, are designed and accommodated in the repository for the purpose of operational convenience and stability after closure. To forecast the change of long-term integrity of the structural components, experimental verification, using in-situ or near in-situ conditions, is necessary. Domestic and foreign requirements with regard to the selection criteria and the performance criteria for structural components in disposal facility were surveyed. Characteristics of various types of cement were studied. Materials and construction methods of structural components similar to those of disposal facility was investigated and test items and methods for integrity of cement concrete were included. Literature survey for domestic groundwater characteristics was performed together with Ca-type bentonite ore which is a potential backfill material. Causes or factors affecting the durability of the cement structures were summarized. Experiments to figure out the ions leaching out from and migrating into cement soaked in distilled water and synthetic groundwater, respectively, were carried out. And finally, diffusion of chloride ion through cement was experimentally measured

Limit load analysis of thick-walled concrete structures

International Nuclear Information System (INIS)

Argyris, J.H.; Faust, G.; Willam, K.J.

1975-01-01

The paper illustrates the interaction of constitutive modeling and finite element solution techniques for limit load prediction of concrete structures. On the constitutive side, an engineering model of concrete fracture is developed in which the Mohr-Coulomb criterion is augmented by tension cut-off to describe incipient failure. Upon intersection with the stress path the failure surface collapses for brittle behaviour according to one of three softening rules, no-tension, no-cohesion, and no-friction. The stress transfer accompanying the energy dissipation during local failure is modelled by several fracture rules which are examined with regard to ultimate load prediction. On the numerical side the effect of finite element idealization is studied first as far as ultimate load convergence is concerned. Subsequently, incremental tangential and initial load techniques are compared together with the effect of step size. Limit load analyses of a thick-walled concrete ring and a lined concrete reactor closure conclude the paper with examples from practical engineering. (orig.) [de

Maxey Flats low-level waste disposal site closure activities

International Nuclear Information System (INIS)

Haight, C.P.; Mills, D.; Razor, J.E.

1987-01-01

The Maxey Flats Radioactive Waste Disposal Facility in Fleming County, Kentucky is in the process of being closed. The facility opened for commercial business in the spring of 1963 and received approximately 4.75 million cubic feet of radioactive waste by the time it was closed in December of 1977. During fourteen years of operation approximately 2.5 million curies of by-product material, 240,000 kilograms of source material, and 430 kilograms of special nuclear material were disposed. The Commonwealth purchased the lease hold estate and rights in May 1978 from the operating company. This action was taken to stabilize the facility and prepare it for closure consisting of passive care and monitoring. To prepare the site for closure, a number of remedial activities had to be performed. The remediation activities implemented have included erosion control, surface drainage modifications, installation of a temporary plastic surface cover, leachate removal, analysis, treatment and evaporation, US DOE funded evaporator concentrates solidification project and their on-site disposal in an improved disposal trench with enhanced cover for use in a humid environment situated in a fractured geology, performance evaluation of a grout injection demonstration, USGS subsurface geologic investigation, development of conceptual closure designs, and finally being added to the US EPA National Priority List for remediation and closure under Superfund. 13 references, 3 figures

A study on sodium-concrete reaction

Energy Technology Data Exchange (ETDEWEB)

Pae, Jae Huem; Min, Byung Hoon; Lee, Joon Sik; Lee, Choong Hui; Chung, Ki Hong; Keum, Choong Ki [Suwon University, Suwon (Korea, Republic of)

1994-07-15

Sodium is commonly used as a coolant in liquid metal reactor. A large amount of its leakage may be possible in hypothetical accidents, even though the possibility is very low. In case that the leaked hot sodium comes in direct contact with structural concrete of liquid metal reactor, the reactor`s integrity can be challenged by the rupture of structure materials, hydrogen generation and its explosion, and release of radioactive aerosols due to sodium-concrete reaction. The knowledge of sodium-concrete reaction is evaluated to be one of the important and indispensable technologies for the establishment of safety measure in liquid metal reactor. In this study, the experimental facility of sodium-concrete reaction is to be designed, constructed and operated. And the reaction phenomena of sodium-concrete reaction is also to be analyzed through the experimental results. The aim of this study is to establish the measure of safety and protection for sodium-related facilities and to secure one of the fundamental technologies of liquid metal reactor safety. 47 refs., 7 figs., 13 tab.

Benchmark experiments of dose distributions in phantom placed behind iron and concrete shields at the TIARA facility

International Nuclear Information System (INIS)

Nakane, Yoshihiro; Sakamoto, Yukio; Tsuda, Shuichi

2004-01-01

To verify the calculation methods used for the evaluations of neutron dose at the radiation shielding design of the high-intensity proton accelerator facility (J-PARC), dose distributions in a plastic phantom of 30x30x30 cm 3 slab placed behind iron and concrete test shields were measured by using a tissue equivalent proportional counter for 65-MeV quasi-monoenergetic neutrons generated from the 7 Li(p,n) reactions with 68-MeV protons at the TIARA facility. Dose distributions in the phantom were calculated by using the MCNPX and the NMTC/JAM-MCNP codes with the flux-to-dose conversion coefficients prepared for the shielding design of the facility. The comparison results show the calculated results were in good agreement with the measured ones within 20%. (author)

Economic analysis of recycling contaminated concrete

Energy Technology Data Exchange (ETDEWEB)

Stephen, A.; Ayers, K.W.; Boren, J.K.; Parker, F.L. [Vanderbilt Univ., Nashville, TN (United States)

1997-02-01

Decontamination and Decommissioning activities in the DOE complex generate large volumes of radioactively contaminated and uncontaminated concrete. Currently, this concrete is usually decontaminated, the contaminated waste is disposed of in a LLW facility and the decontaminated concrete is placed in C&D landfills. A number of alternatives to this practice are available including recycling of the concrete. Cost estimates for six alternatives were developed using a spreadsheet model. The results of this analysis show that recycling alternatives are at least as economical as current practice.

An evaluation of concrete recycling and reuse practices

Energy Technology Data Exchange (ETDEWEB)

Nakhjiri, K.S.; MacKinney, J.

1997-02-01

Nuclear facilities operated by the Department of Energy (DOE), Department of Defense (DOD), and NRC licensees contain many concrete structures that are contaminated with radioactivity. Dismantling these structures will result in significant quantities of waste materials, both contaminated and uncontaminated. Bartlett estimates the total volume of waste from demolition of concrete structures to be on the order of 4 million cubic meters, but that only 20,000 cubic meters would be contaminated with radioactivity. Other studies suggest that as much as 5% of the concrete in these facilities would be contaminated with radioactivity. While the actual quantity of contaminated material should be fixed with greater precision, the fact that so much uncontaminated concrete exists (over 95% of the total 4 million cubic meters) suggests that a program that recycles concrete could produce substantial savings for both government agencies (DOE, DOD) and private companies (NRC licensees). This paper presents a fundamental discussion of (1) various methods of processing concrete, (2) demolition methods, especially those compatible with recycling efforts, and (3) state-of-the-art concrete dismantlement techniques.

Optimizing of the recycling of contaminated concrete debris. Final report

International Nuclear Information System (INIS)

Kloeckner, J.; Rasch, H.; Schloesser, K.H.; Schon, T.

1999-01-01

1. Latest research: So far concrete debris from nuclear facilities has been free released or was treated as radioactive waste. 2. Objective: The objective of this study is to develop solutions and methods for recycling concrete debris. The amount of materials used in nuclear facilities should be limited and the contamination of new materials should be avoided. 3. Methods: The status of recycling was presented using examples of operating or completed decommissioning as well as available studies and literature. The quality requirements for the production of new concrete products using recycled materials has been discussed. The expected amounts of concrete debris for the next 12 years was estimated. For the proposed recycling examples, radiological and economic aspects have been considered. 4. Results: The production of qualified concrete products from concrete debris is possible by using modified receptions. Technical regulations to this are missing. There is no need for the utilization of large amounts of concrete debris for shielding walls. For the production of new shielding-containers for radioactive waste, concrete debris can be applied. Regarding the distance to a central recycling facility the use of mobile equipment can be economical. By using the concrete for filling the cavity or space in a final storage, it is possible to dispose the whole radioactive debris. 5. Application possibilities: The use of concrete debris as an inner concrete shielding in waste-containers today is already possible. For the manufacture of qualified concrete products by using recycling products, further developments and regulations are necessary. (orig.) [de

Construction of concrete hot cells

International Nuclear Information System (INIS)

1981-12-01

The standard is to be applied to rooms (hot cells) which are enclosed by a concrete shield and in which radioactive material is handled by remote control. The rooms may be in facilities for experimental purposes (e.g. development of fuel elements and materials or of chemical processes) or in facilities for production purposes (e.g. reprocessing of nuclear fuel or treatment of radioactive wastes). The standard is to give a design hasis for concrete hot cells and their installations which is to be applied by designers, constructors, future users and competent authorities as well as independent experts. (orig.) [de

Construction of concrete hot cells

International Nuclear Information System (INIS)

1980-09-01

The standard is to be applied to rooms (hot cells) which are enclosed by a concrete shield and in which radioactive material is handled by remote control. The rooms may be in facilities for experimental purposes (e.g. development of fuel elements and materials or of chemical processes) or in facilities for production purposes (e.g. reprocessing of nuclear fuel or treatment of radioactive wastes). The standard is to give a design basis for concrete hot cells and their installations which is to be applied by designers, constructors, future users and competent authorities as well as independent experts. (orig.) [de

Concrete shaver. Innovative technology summary report

International Nuclear Information System (INIS)

1998-12-01

The US Department of Energy (DOE) is in the process of decontamination and decommissioning (D and D) for many of its nuclear facilities throughout the United States. These facilities must be dismantled and the demolition waste sized into manageable pieces for handling and disposal. The facilities undergoing D and D are typically chemically and/or radiologically contaminated. To facilitate this work, DOE requires a tool capable of removing the surface of radiologically contaminated concrete floors. Operating requirements for the tool include simple and economical operation, the capability of operating in ambient temperatures from 3 C to 40 C (37 F to 104 F), and the ability to be easily decontaminated. The tool also must be safe for workers. The Marcrist Industries Limited concrete shaver is an electrically driven, self-propelled concrete and coating removal system. This technology consists of a 25-cm (10-in.)-wide diamond impregnated shaving drum powered by an electric motor and contains a vacuum port for dust extraction. The concrete shaver is ideal for use on open, flat, floor areas. The shaver may also be used on slightly curved surfaces. This shaver is self-propelled and produces a smooth, even surface with little vibration. The concrete shaver is an attractive alternative to traditional pneumatic scabbling tools, which were considered the baseline in this demonstration. The use of this tool reduces worker fatigue (compared to the baseline) due to lower vibration. The shaver is more than five times faster than the five-piston pneumatic scabbler at removing contamination from concrete. Because of this increased productivity, the shaver is 50% less costly to operate than baseline technologies. The DOE has successfully demonstrated the concrete shaver for decontaminating floors for free-release surveys prior to demolition work

The 300 area waste acid treatment system closure plan

International Nuclear Information System (INIS)

Luke, S.N.

1996-01-01

The 300 Area Waste Acid Treatment System (WATS) is located within operable units 300-FF-2 (source) and 300-FF-5 (groundwater), as designated in the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) . Operable units 300-FF-2 and 300-FF-5 are scheduled to be remediated using the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Remedial Investigation/Feasibility Study (RI/FS) process. Thus, any remediation of the 300 Area WATS with respect to contaminants not produced by those facilities and soils and groundwater will be deferred to the CERCLA RI/FS process. Final closure activities will be completed in 3 phases and certified in accordance with the 300 Area WATS closure plan by the Washington State Department of Ecology (Ecology) and the U.S. Environmental Protection Agency (EPA). It is anticipated that the 300 Area WATS closure would take 2 years to complete

Electrokinetic decontamination of concrete

International Nuclear Information System (INIS)

Lomasney, H.L.; SenGupta, A.K.; Yachmenev, V.

1996-01-01

ELECTROSORB Electrokinetic Extraction Technology, developed by ISOTRON Corp., offers a cost-effective approach to treating contaminated concrete. Heavy metals/radionuclides trapped in concrete can be extracted using this process if they are chemically solubilized; solubilizers used are citric acid alone and a mixture of citric and nitric acids. A DC electric field is applied across the contaminated concrete to electrokinetically transport the solubilized contaminants from the concrete pores to a collector on the concrete surface. The collector is an extraction pad laid on the surface. The pad provides confinement for a planar electrode and solubilizer solution; it is operated under a vacuum to hold the pad against the concrete surface. Operation requires little attendance, reducing the workers' health hazards. The process incorporates a mechanism for recycling the solubilizer solution. A field demonstration of the process took place in Building 21 of DOE's Mound facility in Miamisburg, OH, over 12 days in June 1996. The thorium species present in this building's concrete floors included ThO 2 and thorium oxalate. The nitric acid was found to facilitate Th extraction

Closure Report for Corrective Action Unit 547: Miscellaneous Contaminated Waste Sites, Nevada National Security Site, Nevada

Energy Technology Data Exchange (ETDEWEB)

NSTec Environmental Restoration

2012-07-17

This Closure Report (CR) presents information supporting closure of Corrective Action Unit (CAU) 547, Miscellaneous Contaminated Waste Sites, and provides documentation supporting the completed corrective actions and confirmation that closure objectives for CAU 547 were met. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; the U.S. Department of Energy (DOE), Environmental Management; the U.S. Department of Defense; and DOE, Legacy Management (FFACO, 1996 as amended). CAU 547 consists of the following three Corrective Action Sites (CASs), located in Areas 2, 3, and 9 of the Nevada National Security Site: (1) CAS 02-37-02, Gas Sampling Assembly; (2) CAS 03-99-19, Gas Sampling Assembly; AND (3) CAS 09-99-06, Gas Sampling Assembly Closure activities began in August 2011 and were completed in June 2012. Activities were conducted according to the Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) for CAU 547 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2011). The recommended corrective action for the three CASs in CAU 547 was closure in place with administrative controls. The following closure activities were performed: (1) Open holes were filled with concrete; (2) Steel casings were placed over vertical expansion joints and filled with cement; (3) Engineered soil covers were constructed over piping and exposed sections of the gas sampling system components; (4) Fencing, monuments, Jersey barriers, radiological postings, and use restriction (UR) warning signs were installed around the perimeters of the sites; (5) Housekeeping debris was picked up from around the sites and disposed; and (6) Radiological surveys were performed to confirm final radiological postings. UR documentation is included in Appendix D. The post-closure plan was presented in detail in the CADD/CAP for CAU 547 and is included as

Closure Report for Corrective Action Unit 547: Miscellaneous Contaminated Waste Sites, Nevada National Security Site, Nevada

International Nuclear Information System (INIS)

2012-01-01

This Closure Report (CR) presents information supporting closure of Corrective Action Unit (CAU) 547, Miscellaneous Contaminated Waste Sites, and provides documentation supporting the completed corrective actions and confirmation that closure objectives for CAU 547 were met. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; the U.S. Department of Energy (DOE), Environmental Management; the U.S. Department of Defense; and DOE, Legacy Management (FFACO, 1996 as amended). CAU 547 consists of the following three Corrective Action Sites (CASs), located in Areas 2, 3, and 9 of the Nevada National Security Site: (1) CAS 02-37-02, Gas Sampling Assembly; (2) CAS 03-99-19, Gas Sampling Assembly; AND (3) CAS 09-99-06, Gas Sampling Assembly Closure activities began in August 2011 and were completed in June 2012. Activities were conducted according to the Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) for CAU 547 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2011). The recommended corrective action for the three CASs in CAU 547 was closure in place with administrative controls. The following closure activities were performed: (1) Open holes were filled with concrete; (2) Steel casings were placed over vertical expansion joints and filled with cement; (3) Engineered soil covers were constructed over piping and exposed sections of the gas sampling system components; (4) Fencing, monuments, Jersey barriers, radiological postings, and use restriction (UR) warning signs were installed around the perimeters of the sites; (5) Housekeeping debris was picked up from around the sites and disposed; and (6) Radiological surveys were performed to confirm final radiological postings. UR documentation is included in Appendix D. The post-closure plan was presented in detail in the CADD/CAP for CAU 547 and is included as

Test installation for separation of contaminated/activated concrete

International Nuclear Information System (INIS)

Klok, H.; Peeze Binkhorst, I.A.G.M.

1993-01-01

Large amounts of contaminated concrete are released when dismantling nuclear power plants. Volume reduction of contaminated concrete has economic and environmental advantages. A test facility has been built by which contaminated concrete can be separated in clean aggregate and contaminated cement. During the tests a considerable amount of dust accumulated in the valve construction of the test installation. This could result in an unacceptable accumulation of contaminated dust, blocking of the handle and emission of dust and grit during the operation of the test facility. The design of the valve construction was adjusted by using and testing flexible materials

Influence of processing factors over concrete strength.

Science.gov (United States)

Kara, K. A.; Dolzhenko, A. V.; Zharikov, I. S.

2018-03-01

Construction of facilities of cast in-situ reinforced concrete poses additional requirements to quality of material, peculiarities of the construction process may sometimes lead to appearance of lamination planes and inhomogeneity of concrete, which reduce strength of the material and structure as a whole. Technology compliance while working with cast in-situ concrete has a significant impact onto the concrete strength. Such process factors as concrete curing, vibration and compaction of the concrete mixture, temperature treatment, etc., when they are countered or inadequately followed lead to a significant reduction in concrete strength. Here, the authors experimentally quantitatively determine the loss of strength in in-situ cast concrete structures due to inadequate following of process requirements, in comparison with full compliance.

Conceptual model for concrete long time degradation in a deep nuclear waste repository

Energy Technology Data Exchange (ETDEWEB)

Lagerblad, B; Traegaardh, J [Swedish Cement and Concrete Research Inst., Stockholm (Sweden)

1994-02-01

This report is mainly a state-of-the-art report of concrete long time durability in the environment expected in a deep site underground nuclear waste repository in Swedish crystalline bedrock. The report treats how the concrete and the surrounding groundwater will interact and how they will be affected by cement chemistry, type of aggregate etc. The different mechanisms for concrete alteration treated include sulphate attack, carbonation, chloride attack, alkali-silica reaction and leaching phenomena. In a long time perspective, the chemical alterations in concrete is mainly governed by the surrounding groundwater composition. After closure the composition of the groundwater will change character from a modified meteoric to a saline composition. Therefore two different simulated groundwater compositions have been used in modelling the chemical interaction between concrete and groundwater. The report also includes a study of old and historical concrete which show observations concerning recrystallization phenomena in concrete. 72 refs, 39 figs.

Conceptual model for concrete long time degradation in a deep nuclear waste repository

International Nuclear Information System (INIS)

Lagerblad, B.; Traegaardh, J.

1994-02-01

This report is mainly a state-of-the-art report of concrete long time durability in the environment expected in a deep site underground nuclear waste repository in Swedish crystalline bedrock. The report treats how the concrete and the surrounding groundwater will interact and how they will be affected by cement chemistry, type of aggregate etc. The different mechanisms for concrete alteration treated include sulphate attack, carbonation, chloride attack, alkali-silica reaction and leaching phenomena. In a long time perspective, the chemical alterations in concrete is mainly governed by the surrounding groundwater composition. After closure the composition of the groundwater will change character from a modified meteoric to a saline composition. Therefore two different simulated groundwater compositions have been used in modelling the chemical interaction between concrete and groundwater. The report also includes a study of old and historical concrete which show observations concerning recrystallization phenomena in concrete. 72 refs, 39 figs

Post-Closure Inspection Report for Corrective Action Unit 92: Area 6 Decon Pond Facility, Nevada Test Site, Nevada, for Calendar Year 2006

International Nuclear Information System (INIS)

NSTec Environmental Restoration

2007-01-01

This Post-Closure Inspection Report provides an analysis and summary of inspections for Corrective Action Unit (CAU) 92, Area 6 Decon Pond Facility. CAU 92 was closed according to the ''Resource Conservation and Recovery Act'' (RCRA) Part B Operational Permit (Nevada Division of Environmental Protection [NDEP], 1995) and the ''Federal Facility Agreement and Consent Order'' (FFACO) of 1996 (FFACO, 1996). Closure activities were completed on February 16, 1999, and the Closure Report (U.S. Department of Energy, Nevada Operations Office, 1999) was approved and a Notice of Completion issued by NDEP on May 11, 1999. CAU 92 consists of two Corrective Action Sites (CASs), CAS 06-04-01, Decon Pad Oil/Water Separator; and CAS 06-05-02, Decontamination Pond (RCRA). Both CASs have use restrictions; however, only CAS 06-05-02 requires post-closure inspections. Visual inspections of the cover and fencing at CAS 06-05-02 are performed quarterly. Additional inspections are conducted if precipitation occurs in excess of 1.28 centimeters (cm) (0.50 inches [in.]) in a 24-hour period. This report covers calendar year 2006. Quarterly site inspections were performed in March, June, September, and December of 2006. All observations indicated the continued integrity of the unit. No issues or concerns were noted, and no corrective actions were necessary. Copies of the inspection checklists and field notes completed during each inspection are included in Appendix A of this report, and photographs taken during the site inspections are included in Appendix B of this report. One additional inspection was performed after a precipitation event that exceeded 1.28 cm (0.50 in.) within a 24-hour period during 2006. No significant changes in site conditions were noted during this inspection, and no corrective actions were necessary. A copy of the inspection checklist and field notes completed during this additional inspection is included in Appendix A of this report. Precipitation records for 2006

Hanford Patrol Academy demolition sites closure plan

Energy Technology Data Exchange (ETDEWEB)

1993-09-30

The Hanford Site is owned by the U.S. Government and operated by the U.S. Department of Energy, Richland Operations Office. Westinghouse Hanford Company is a major contractor to the U.S. Department of Energy, Richland Operations Office and serves as co-operator of the Hanford Patrol Academy Demolition Sites, the unit addressed in this paper. This document consists of a Hanford Facility Dangerous Waste Part A Permit Application, Form 3 (Revision 4), and a closure plan for the site. An explanation of the Part A Form 3 submitted with this closure plan is provided at the beginning of the Part A section. This Hanford Patrol Academy Demolition Sites Closure Plan submittal contains information current as of December 15, 1994.

Mobilization plan for the Y-12 9409-5 tank storage facility RCRA closure plan. Final report. Revision 1

International Nuclear Information System (INIS)

1993-11-01

This mobilization plan identifies the activities and equipment necessary to begin the field sampling for the Oak Ridge Y-12 9409-5 Diked Tank Storage Facility (DTSF) Resource Conservation and Recovery Act (RCRA) closure. Elements of the plan outline the necessary components of each mobilization task and identify whether SAIC or the Martin Marietta Energy Systems, Inc. Y-12 Environmental Restoration Division will be responsible for task coordination. Field work will be conducted in two phases: mobilization phase and soil sampling phase. Training and medical monitoring, access, permits and passes, decontamination/staging area, equipment, and management are covered in this document

Performance of Waterless Concrete

Science.gov (United States)

Toutanji, Houssam; Evans, Steve; Grugel, Richard N.

2010-01-01

The development of permanent lunar bases is constrained by performance of construction materials and availability of in-situ resources. Concrete seems a suitable construction material for the lunar environment, but water, one of its major components, is an extremely scarce resource on the Moon. This study explores an alternative to hydraulic concrete by replacing the binding mix of concrete (cement and water) with sulfur. Sulfur is a volatile element on the lunar surface that can be extracted from lunar soils by heating. Sulfur concrete mixes were prepared to investigate the effect of extreme environmental conditions on the properties of sulfur concrete. A hypervelocity impact test was conducted, having as its target a 5-cm cubic sample of sulfur concrete. This item consisted of JSC-1 lunar regolith simulant (65%) and sulfur (35%). The sample was placed in the MSFC Impact Test Facility s Micro Light Gas Gun target chamber, and was struck by a 1-mm diameter (1.4e-03 g) aluminum projectile at 5.85 km/s. In addition, HZTERN code, provided by NASA was used to study the effectiveness of sulfur concrete when subjected to space radiation.

Offshore concrete structures

International Nuclear Information System (INIS)

Lamas Pardo, M.; Carral Couce, L. M.

2011-01-01

proof that in certain cases the benefits to the steel tip the lance toward the concrete. concrete Gravity Baseplaforms type Con deep have been building since the first unit became operational in 1976, together with a few floating platforms, that in its geometry are similar to those made of steel (barges, semi-submersibles and TLP type platforms). Some of the concepts in concrete that are being emerging last years are floating and gravity base platforms for use as offshore LNG terminals and as offshore industrial plants. The life of these barges can be designed up to 200 years, as the floating Nkossabarge, so they con be a good alternative to the construction of these facilities an land, thus avoiding landfills on the coast that degrade the already punished coast in industrialized countries. the challenge is precisely to optimize their capital costs to complete for an offshore installation against a shore facility. The environment will undoubtedly benefit from this great challenge that lies ahead in the XXI century. (Author) 29 refs.

The Necessary and Sufficient Closure Process Completion Report for Purex FacilitySurveillance and Maintenance

International Nuclear Information System (INIS)

Gerald, J.W.

1997-10-01

This document completes the U.S. Department of Energy Closure Process for Necessary and Sufficient Sets of Standards process for the Plutonium Uranium Extraction facility located at the Hanford Site in Washington State. This documentation is provided to support the Work Smart Standards set identified for the long-term surveillance and maintenance of PUREX. This report is organized into two volumes. Volume 1 contains the following sections: Section 1: Provides an introduction for the document Section 2: Provides a basis for initiating the N ampersand S process Section 3: Defines the work and hazards to be addressed Section 4: Identifies the N ampersand S set of standards and requirements Section 5: Provides the justification for adequacy of the work smart standards Section 6: Shows the criteria and qualifications of the teams Section 7: Describes the stakeholder participation and concerns Section 8: Provides a list of references used within the document

Biodecontamination of concrete surfaces: Occupational and environmental benefits

International Nuclear Information System (INIS)

Johnson, L.; Rogers, R.; Hamilton, M.; Nelson, L.

1996-01-01

Managers and engineers around the globe are presently challenged by high estimated costs for the decontamination and decommissioning of nuclear facilities which are no longer needed or are abandoned. It has been estimated that more than 73 Km 2 of contaminated concrete currently exists in the USDOE complex and is increased many fold when similar facilities are accounted for in other countries. Needs for the decontamination of concrete have been identified as: more cost effective decontamination methods, reduction of secondary wastes, minimized worker radiation exposures and, contaminant containment. Recently environmental microbes have been harnessed to remove the surface of concrete as a technique for decontamination and decommissioning (D and D). This biodecontamination technology has been tested by INEL and BNFL scientists and engineers in both US and United Kingdom nuclear facilities. Biodecontamination field tests at a shutdown nuclear reactor facility (EBR-I) have shown radioactively contaminated surface removed at rates of 4--8 mm/yr, thus validating the feasibility of this technology. Engineering economic analyses indicate two attractive benefits embedded in this approach to concrete D and D: (1) due to the passive nature of the technique, a cost savings of more than an order of magnitude is projected compared to the current labor intensive physical decontamination techniques; and (2) the exposure to humans and the natural environment is greatly reduced due to the unattended, highly contained biodecontamination process

Closure Plan for the Area 5 Radioactive Waste Management Site at the Nevada Test Site

International Nuclear Information System (INIS)

NSTec Environmental Management

2008-01-01

The Area 5 Radioactive Waste Management Site (RMWS) at the Nevada Test Site (NTS) is managed and operated by National Security Technologies, LLC (NSTec), for the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO). This document is the first update of the preliminary closure plan for the Area 5 RWMS at the NTS that was presented in the Integrated Closure and Monitoring Plan (DOE, 2005a). The major updates to the plan include a new closure schedule, updated closure inventory, updated site and facility characterization data, the Title II engineering cover design, and the closure process for the 92-Acre Area of the RWMS. The format and content of this site-specific plan follows the Format and Content Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Closure Plans (DOE, 1999a). This interim closure plan meets closure and post-closure monitoring requirements of the order DOE O 435.1, manual DOE M 435.1-1, Title 40 Code of Federal Regulations (CFR) Part 191, 40 CFR 265, Nevada Administrative Code (NAC) 444.743, and Resource Conservation and Recovery Act (RCRA) requirements as incorporated into NAC 444.8632. The Area 5 RWMS accepts primarily packaged low-level waste (LLW), low-level mixed waste (LLMW), and asbestiform low-level waste (ALLW) for disposal in excavated disposal cells

Closure Plan for the Area 5 Radioactive Waste Management Site at the Nevada Test Site

Energy Technology Data Exchange (ETDEWEB)

NSTec Environmental Management

2008-09-01

The Area 5 Radioactive Waste Management Site (RMWS) at the Nevada Test Site (NTS) is managed and operated by National Security Technologies, LLC (NSTec), for the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO). This document is the first update of the preliminary closure plan for the Area 5 RWMS at the NTS that was presented in the Integrated Closure and Monitoring Plan (DOE, 2005a). The major updates to the plan include a new closure schedule, updated closure inventory, updated site and facility characterization data, the Title II engineering cover design, and the closure process for the 92-Acre Area of the RWMS. The format and content of this site-specific plan follows the Format and Content Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Closure Plans (DOE, 1999a). This interim closure plan meets closure and post-closure monitoring requirements of the order DOE O 435.1, manual DOE M 435.1-1, Title 40 Code of Federal Regulations (CFR) Part 191, 40 CFR 265, Nevada Administrative Code (NAC) 444.743, and Resource Conservation and Recovery Act (RCRA) requirements as incorporated into NAC 444.8632. The Area 5 RWMS accepts primarily packaged low-level waste (LLW), low-level mixed waste (LLMW), and asbestiform low-level waste (ALLW) for disposal in excavated disposal cells.

Surface concrete decontamination equipment developed by Pacific Northwest Laboratory

International Nuclear Information System (INIS)

Halter, J.M.; Sullivan, R.G.; Bevan, J.L.

1982-08-01

This report documents a project that the Pacific Northwest Laboratory conducted to identify and develop techniques for removing contaminated concrete surfaces. A major problem associated with nuclear facility decontamination and decommissioning is how to economically demolish and dispose of contaminated concrete. Removing only the contaminated portion of the concrete can substantially reduce costs. Evaluation of various methods for removing concrete surfaces shows that several techniques presently used require excessive manpower, time, and energy. Many times more material is removed than necessary, increasing the quantity of waste that must be handled under controlled conditions. These evaluations generated the basic criteria for developing a suitable concrete removal technique: provide a convenient method for cleaning surfaces (such as those contaminated by a small spill); reduce the contaminated waste volume that has to be placed into controlled storage; remove surfaces quickly; and minimize personal exposure to potentially harmful radiation or toxic materials. Removal to 1/4 to 1/2 in. of contaminated surface layer is sufficient for cleanup of most facilities. Two unique decontamination methods have been developed: the concrete spaller and the water cannon. The concrete spaller is the most efficient technique: it removes the concrete surface faster than the water cannons and at a lower cost (as little as $3.00/ft 2 of concrete surface). However, the .458 magnum water cannon may be well suited for small or hard-to-reach locations

Successful completion of a RCRA closure for the Fernald Environmental Management Project

International Nuclear Information System (INIS)

Lippitt, J.M.; Kolthoff, K.

1995-01-01

This paper discusses the successful completion of a RCRA (Resource Conservation and Recovery Act) closure of a HF (hydrofluoric acid) tank car at FEMP, which is on the national priorities list of hazardous waste sites and is undergoing CERCLA remediation. The HF tank car closure was conducted by FERMCO. Through a combination of sound planning and team work, the HF tank car was closed safely and ahead of schedule. During > 22,000 hr field work required for construction modifications and neutralization of 9,600 gallons of HF and decontamination rinseates, there were no OSHA recordable incidents. The system design avoided additional costs by maximizing use of existing equipment and facilities. This successful closure of the HF tank car demonstrates FEMP's commitment to reducing risks and cleaning up the facility in a manner consistent with objectives of RCRA regulations and the Ohio EPA hazardous waste rules. This in turn facilitated ongoing negotiations with Ohio EPA to integrate RCRA closure and the ongoing CERCLA remediation activities. This paper addresses why the unit was clean closed under an approved RCRA Closure Plan. Integration of EPA regulations for RCRA and CERCLA programs and the DOE-Orders impacting design, construction and operation of an acid neutralization system is also reviewed. The paper concludes with a discussion of lessons learned in the process in preparing the closure plant and through final project close out

Requirements for thermal insulation on prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Neylan, A.J.; Wistrom, J.D.

1979-01-01

During the past decade, extensive design, construction, and operating experience on concrete pressure vessels for gas-cooled reactor applications has accumulated. Excellent experience has been obtained to date on the structural components (concrete, prestressing systems, liners, penetrations, and closures) and the thermal insulation. Three fundamentally different types of insulation systems have been employed to ensure the satisfactory performance of this component, which is critical to the overall success of the prestressed concrete reactor vessel (PCRV). Although general design criteria have been published, the requirements for design, materials, and construction are not rigorously addressed in any national or international code. With the more onerous design conditions being imposed by advanced reactor systems, much greater attention has been directed to advance the state of the art of insulation systems for PCRVs. This paper addresses some of the more recent developments in this field being performed by General Atomic Company and others. (author)

Criticality assessment of LLRWDF closure

International Nuclear Information System (INIS)

Sarrack, A.G.; Weber, J.H.; Woody, N.D.

1992-01-01

During the operation of the Low Level Radioactive Waste Disposal Facility (LLRWDF), large amounts (greater than 100 kg) of enriched uranium (EU) were buried. This EU came primarily from the closing and decontamination of the Naval Fuels Facility in the time period from 1987--1989. Waste Management Operations (WMO) procedures were used to keep the EU boxes separated to prevent possible criticality during normal operation. Closure of the LLRWDF is currently being planned, and waste stabilization by Dynamic Compaction (DC) is proposed. Dynamic compaction will crush the containers in the LLRWDF and result in changes in their geometry. Research of the LLRWDF operations and record keeping practices have shown that the EU contents of trenches are known, but details of the arrangement of the contents cannot be proven. Reviews of the trench contents, combined with analysis of potential critical configurations, revealed that some portions of the LLRWDF can be expected to be free of criticality concerns while other sections have credible probabilities for the assembly of a critical mass, even in the uncompacted configuration. This will have an impact on the closure options and which trenches can be compacted

Closure Report for Corrective Action Unit 224: Decon Pad and Septic Systems, Nevada Test Site, Nevada

International Nuclear Information System (INIS)

NSTec Environmental Restoration

2007-01-01

Corrective Action Unit (CAU) 224 is located in Areas 02, 03, 05, 06, 11, and 23 of the Nevada Test Site, which is situated approximately 65 miles northwest of Las Vegas, Nevada. CAU 224 is listed in the Federal Facility Agreement and Consent Order (FFACO) of 1996 as Decon Pad and Septic Systems and is comprised of the following nine Corrective Action Sites (CASs): CAS 02-04-01, Septic Tank (Buried); CAS 03-05-01, Leachfield; CAS 05-04-01, Septic Tanks (4)/Discharge Area; CAS 06-03-01, Sewage Lagoons (3); CAS 06-05-01, Leachfield; CAS 06-17-04, Decon Pad and Wastewater Catch; CAS 06-23-01, Decon Pad Discharge Piping; CAS 11-04-01, Sewage Lagoon; and CAS 23-05-02, Leachfield. The Nevada Division of Environmental Protection (NDEP)-approved corrective action alternative for CASs 02-04-01, 03-05-01, 06-03-01, 11-04-01, and 23-05-02 is no further action. As a best management practice, the septic tanks and distribution box were removed from CASs 02-04-01 and 11-04-01 and disposed of as hydrocarbon waste. The NDEP-approved correction action alternative for CASs 05-04-01, 06-05-01, 06-17-04, and 06-23-01 is clean closure. Closure activities for these CASs included removing and disposing of radiologically and pesticide-impacted soil and debris. CAU 224 was closed in accordance with the NDEP-approved CAU 224 Corrective Action Plan (CAP). The closure activities specified in the CAP were based on the recommendations presented in the CAU 224 Corrective Action Decision Document (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, 2005). This Closure Report documents CAU 224 closure activities. During closure activities, approximately 60 cubic yards (yd3) of mixed waste in the form of soil and debris; approximately 70 yd 3 of sanitary waste in the form of soil, liquid from septic tanks, and concrete debris; approximately 10 yd 3 of hazardous waste in the form of pesticide-impacted soil; approximately 0.5 yd 3 of universal waste in the form of

TRANSPORT THROUGH CRACKED CONCRETE: LITERATURE REVIEW

Energy Technology Data Exchange (ETDEWEB)

Langton, C.

2012-05-11

Concrete containment structures and cement-based fills and waste forms are used at the Savannah River Site to enhance the performance of shallow land disposal systems designed for containment of low-level radioactive waste. Understanding and measuring transport through cracked concrete is important for describing the initial condition of radioactive waste containment structures at the Savannah River Site (SRS) and for predicting performance of these structures over time. This report transmits the results of a literature review on transport through cracked concrete which was performed by Professor Jason Weiss, Purdue University per SRR0000678 (RFP-RQ00001029-WY). This review complements the NRC-sponsored literature review and assessment of factors relevant to performance of grouted systems for radioactive waste disposal. This review was performed by The Center for Nuclear Waste Regulatory Analyses, San Antonio, TX, and The University of Aberdeen, Aberdeen Scotland and was focused on tank closure. The objective of the literature review on transport through cracked concrete was to identify information in the open literature which can be applied to SRS transport models for cementitious containment structures, fills, and waste forms. In addition, the literature review was intended to: (1) Provide a framework for describing and classifying cracks in containment structures and cementitious materials used in radioactive waste disposal, (2) Document the state of knowledge and research related to transport through cracks in concrete for various exposure conditions, (3) Provide information or methodology for answering several specific questions related to cracking and transport in concrete, and (4) Provide information that can be used to design experiments on transport through cracked samples and actual structures.

Transport Through Cracked Concrete: Literature Review

International Nuclear Information System (INIS)

Langton, C.

2012-01-01

Concrete containment structures and cement-based fills and waste forms are used at the Savannah River Site to enhance the performance of shallow land disposal systems designed for containment of low-level radioactive waste. Understanding and measuring transport through cracked concrete is important for describing the initial condition of radioactive waste containment structures at the Savannah River Site (SRS) and for predicting performance of these structures over time. This report transmits the results of a literature review on transport through cracked concrete which was performed by Professor Jason Weiss, Purdue University per SRR0000678 (RFP-RQ00001029-WY). This review complements the NRC-sponsored literature review and assessment of factors relevant to performance of grouted systems for radioactive waste disposal. This review was performed by The Center for Nuclear Waste Regulatory Analyses, San Antonio, TX, and The University of Aberdeen, Aberdeen Scotland and was focused on tank closure. The objective of the literature review on transport through cracked concrete was to identify information in the open literature which can be applied to SRS transport models for cementitious containment structures, fills, and waste forms. In addition, the literature review was intended to: (1) Provide a framework for describing and classifying cracks in containment structures and cementitious materials used in radioactive waste disposal, (2) Document the state of knowledge and research related to transport through cracks in concrete for various exposure conditions, (3) Provide information or methodology for answering several specific questions related to cracking and transport in concrete, and (4) Provide information that can be used to design experiments on transport through cracked samples and actual structures.

TECHNICAL PEER REVIEW REPORT - YUCCA MOUNTAIN: WASTE PACKAGE CLOSURE CONTROL SYSTEM

Energy Technology Data Exchange (ETDEWEB)

NA

2005-10-25

The objective of the Waste Package Closure System (WPCS) project is to assist in the disposal of spent nuclear fuel (SNF) and associated high-level wastes (HLW) at the Yucca Mountain site in Nevada. Materials will be transferred from the casks into a waste package (WP), sealed, and placed into the underground facility. The SNF/HLW transfer and closure operations will be performed in an aboveground facility. The objective of the Control System is to bring together major components of the entire WPCS ensuring that unit operations correctly receive, and respond to, commands and requests for data. Integrated control systems will be provided to ensure that all operations can be performed remotely. Maintenance on equipment may be done using hands-on or remote methods, depending on complexity, exposure, and ease of access. Operating parameters and nondestructive examination results will be collected and stored as permanent electronic records. Minor weld repairs must be performed within the closure cell if the welds do not meet the inspection acceptance requirements. Any WP with extensive weld defects that require lids to be removed will be moved to the remediation facility for repair.

TECHNICAL PEER REVIEW REPORT - YUCCA MOUNTAIN: WASTE PACKAGE CLOSURE CONTROL SYSTEM

International Nuclear Information System (INIS)

2005-01-01

The objective of the Waste Package Closure System (WPCS) project is to assist in the disposal of spent nuclear fuel (SNF) and associated high-level wastes (HLW) at the Yucca Mountain site in Nevada. Materials will be transferred from the casks into a waste package (WP), sealed, and placed into the underground facility. The SNF/HLW transfer and closure operations will be performed in an aboveground facility. The objective of the Control System is to bring together major components of the entire WPCS ensuring that unit operations correctly receive, and respond to, commands and requests for data. Integrated control systems will be provided to ensure that all operations can be performed remotely. Maintenance on equipment may be done using hands-on or remote methods, depending on complexity, exposure, and ease of access. Operating parameters and nondestructive examination results will be collected and stored as permanent electronic records. Minor weld repairs must be performed within the closure cell if the welds do not meet the inspection acceptance requirements. Any WP with extensive weld defects that require lids to be removed will be moved to the remediation facility for repair

Interim safety basis for fuel supply shutdown facility

International Nuclear Information System (INIS)

Brehm, J.R.; Deobald, T.L.; Benecke, M.W.; Remaize, J.A.

1995-01-01

This ISB in conjunction with the new TSRs, will provide the required basis for interim operation or restrictions on interim operations and administrative controls for the Facility until a SAR is prepared in accordance with the new requirements. It is concluded that the risk associated with the current operational mode of the Facility, uranium closure, clean up, and transition activities required for permanent closure, are within Risk Acceptance Guidelines. The Facility is classified as a Moderate Hazard Facility because of the potential for an unmitigated fire associated with the uranium storage buildings

Forterra Concrete Products, Inc.

Science.gov (United States)

The EPA is providing notice of a proposed Administrative Penalty Assessment against Forterra Concrete Products, Inc., a business located at 511 E. John Carpenter Freeway, Irving, TX, 75062, for alleged violations at its facility located at 23600 W. 40th St

Effect of High Temperature or fire on heavy weight concrete properties used in nuclear facilities

International Nuclear Information System (INIS)

Sakr, K.

2003-01-01

In the present work the effect of different duration (1, 2 and 3 hours) of high temperatures (250 degree C, 500 degree C, 750 degree C and 950 degree C) on the physical and mechanical properties of heavy concrete shields were studied. The effect of fire fitting systems on ordinary concrete was investigated. The work was extended to determine the effect of high temperature or accidental fire on the radiation properties of heavy weight concrete. Results showed that ilmenite concrete had the highest density, absorption, and modulus of elasticity when compared to the other types of studied concrete and it had also higher values of compressive, tensile, bending and bonding strength than ordinary or baryte concrete. Ilmenite concrete had the highest attenuation of transmitted gamma rays in comparing to gravel concrete and baryte concrete. Ilmenite concrete was more resistant to elevated temperature than gravel concrete and baryte concrete. Foam or air as a fire fitting system in concrete structure that exposed to high temperature or accidental fire proved that better than water

Structural Materials: 95. Concrete

International Nuclear Information System (INIS)

Naus, Dan J.

2012-01-01

Nuclear power plant concrete structures and their materials of construction are described, and their operating experience noted. Aging and environmental factors that can affect the durability of the concrete structures are identified. Basic components of a program to manage aging of these structures are identified and described. Application of structural reliability theory to devise uniform risk-based criteria by which existing facilities can be evaluated to achieve a desired performance level when subjected to uncertain demands and to quantify the effects of degradation is outlined. Finally, several areas are identified where additional research is desired.

Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

International Nuclear Information System (INIS)

1975-12-01

The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

A new system for crack closure of cementitious materials using shrinkable polymers

International Nuclear Information System (INIS)

Jefferson, Anthony; Joseph, Christopher; Lark, Robert; Isaacs, Ben; Dunn, Simon; Weager, Brendon

2010-01-01

This paper presents details of an original crack-closure system for cementitious materials using shrinkable polymer tendons. The system involves the incorporation of unbonded pre-oriented polymer tendons in cementitious beams. Crack closure is achieved by thermally activating the shrinkage mechanism of the restrained polymer tendons after the cement-based material has undergone initial curing. The feasibility of the system is demonstrated in a series of small scale experiments on pre-cracked prismatic mortar specimens. The results from these tests show that, upon activation, the polymer tendon completely closes the preformed macro-cracks and imparts a significant stress across the crack faces. The potential of the system to enhance the natural autogenous crack healing process and generally improve the durability of concrete structures is addressed.

Issues related to the licensing of final disposal facilities for radioactive waste

International Nuclear Information System (INIS)

Medici, M.A.; Alvarez, D.E.; Lee Gonzales, H.; Piumetti, E.H.; Palacios, E.

2010-01-01

The licensing process of a final disposal facility for radioactive waste involves the design, construction, pre-operation, operation, closure and post closure stages. While design and pre-operational stages are, to a reasonable extent, similar to other kind of nuclear or radioactive facilities, construction, operation, closure and post-closure of a radioactive waste disposal facility have unique meanings. As consequence of that, the licensing process should incorporate these particularities. Considering the long timeframes involved at each stage of a waste disposal facility, it is convenient that the development of the project being implemented in and step by step process, be flexible enough as to adapt to new requirements that would arise as a consequence of technology improvements or due to variations in the socio-economical and political conditions. In Argentina, the regulatory Standard AR 0.1.1 establishes the general guideline for the 'Licensing of Class I facilities (relevant facilities)'. Nevertheless, for radioactive waste final disposal facilities a new specific guidance should be developed in addition to the Basic Standard mentioned. This paper describes the particularities of final disposal facilities indicating that a specific licensing system for this type of facilities should be foreseen. (authors) [es

Revision of 'JASS 5N reinforced concrete work for nuclear power facilities'

International Nuclear Information System (INIS)

Masuda, Yoshihiro; Kitagawa, Takashi

2013-01-01

'JASS 5N, Reinforced Concrete Work at Nuclear Power Plants,' is part of the 'Japanese Architectural Standard Specification and Its Interpretation' established by the Architectural Institute of Japan. It is the stipulation to establish the standards for the implementation of reinforced concrete work and quality control for the major buildings of nuclear power plants, and to ensure the safety related to the construction work. The original specification was established in 1985, and its third revised edition was published in February 2013. This 2013 edition is composed of 15 sections and four items of appendices. This paper introduces the major revisions of each section, and explains the newly added section 'Section 14: Small-scale Reinforced Concrete Work.' In addition, this paper describes the newly added 'Appendix: Quality Standards for Heavy Mortal (tentative draft),' and the minor change that part of the appendix related to reinforced concrete was taken into the interpretation of 'Section 10: Reinforced Concrete Work.' (O.A.)

Summary of Uranium Solubility Studies in Concrete Waste Forms and Vadose Zone Environments

Energy Technology Data Exchange (ETDEWEB)

Golovich, Elizabeth C.; Wellman, Dawn M.; Serne, R. Jeffrey; Bovaird, Chase C.

2011-09-30

One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Concrete encasement would contain and isolate the waste packages from the hydrologic environment and act as an intrusion barrier. The current plan for waste isolation consists of stacking low-level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages in concrete. These concrete-encased waste stacks are expected to vary in size with maximum dimensions of 6.4 m long, 2.7 m wide, and 4 m high. The waste stacks are expected to have a surrounding minimum thickness of 15 cm of concrete encasement. These concrete-encased waste packages are expected to withstand environmental exposure (solar radiation, temperature variations, and precipitation) until an interim soil cover or permanent closure cover is installed and to remain largely intact thereafter. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. This report presents the results of investigations elucidating the uranium mineral phases controlling the long-term fate of uranium within concrete waste forms and the solubility of these phases in concrete pore waters and alkaline, circum-neutral vadose zone environments.

Defense Waste Processing Facility Canister Closure Weld Current Validation Testing

Energy Technology Data Exchange (ETDEWEB)

Korinko, P. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Maxwell, D. N. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2018-01-29

Two closure welds on filled Defense Waste Processing Facility (DWPF) canisters failed to be within the acceptance criteria in the DWPF operating procedure SW4-15.80-2.3 (1). In one case, the weld heat setting was inadvertently provided to the canister at the value used for test welds (i.e., 72%) and this oversight produced a weld at a current of nominally 210 kA compared to the operating procedure range (i.e., 82%) of 240 kA to 263 kA. The second weld appeared to experience an instrumentation and data acquisition upset. The current for this weld was reported as 191 kA. Review of the data from the Data Acquisition System (DAS) indicated that three of the four current legs were reading the expected values, approximately 62 kA each, and the fourth leg read zero current. Since there is no feasible way by further examination of the process data to ascertain if this weld was actually welded at either the target current or the lower current, a test plan was executed to provide assurance that these Nonconforming Welds (NCWs) meet the requirements for strength and leak tightness. Acceptance of the welds is based on evaluation of Test Nozzle Welds (TNW) made specifically for comparison. The TNW were nondestructively and destructively evaluated for plug height, heat tint, ultrasonic testing (UT) for bond length and ultrasonic volumetric examination for weld defects, burst pressure, fractography, and metallography. The testing was conducted in agreement with a Task Technical and Quality Assurance Plan (TTQAP) (2) and applicable procedures.

Separation of contaminated concrete

International Nuclear Information System (INIS)

Bakiewicz, J.L.; Reymer, A.P.S.

1990-01-01

Separating the contaminated parts from the non-contaminated parts from decommissioned nuclear facilities may strongly reduce the amount of contaminated concrete. The reduction in volume of the radioactive contaminated concrete is dependent on how much cementstone is in the concrete. This research program shows that the radioactive contamination is mostly in the cementstone. However the choice that the cementstone parts, (or better said the radioactive parts) are smaller than 1 mm may not always be true. Normally the cementstone takes about 30% of the total concrete volume. A separation procedure composed by a combination of milling and thermal shock has been assessed. Both the cold and hot thermal shock in combination with milling are not able to separate the cementstone from the larger aggregates completely. However, the cementstone from the concrete with a low nominal grain size seems to be almost completely removed by the combination cold thermal shock/milling, while the cementstone from the concrete with a high nominal grain size seems to be almost completely removed by the combination hot thermal shock/milling. After both methods a layer of cementstone was still visible on the aggregates. Washing followed by a nitric acid treatment removed each 2 wt% of cementstone

Washing-electrokinetic Decontamination for Concrete Contaminated with Cobalt and Cesium

International Nuclear Information System (INIS)

Kim, Gye Nam; Yang, Byeong Il; Choi, Wang Kyu; Lee, Kune Woo; Hyeon, Jay Hyeok

2009-01-01

A great volume of radioactive concrete is generated during the operation and the decommissioning of nuclear facilities. The washing-electrokinetic technology in this study, which combined an electrokinetic method and a washing method, was developed to decontaminate the concrete generated in nuclear facilities. The results of only an electrokinetic decontamination for the concrete showed that cobalt was removed to below 1% from the concrete due to its high pH. Therefore, the washing electrokinetic technology was applied to lower the pH of the concrete. Namely, when the concrete was washed with 3 M of hydrochloric acid for 4 hours (0.17 day), the CaCO 3 in the concrete was decomposed into CO 2 and the pH of the concrete was reduced to 3.7, and the cobalt and cesium in the concrete were removed by up to 85.0% and 76.3% respectively. Next, when the washed concrete was decontaminated by the electrokinetic method with 0.01M of acetic acid in the 1L electrokinetic equipment for 14.83 days, the cobalt and the cesium in the concrete were both removed by up to 99.7% and 99.6% respectively. The removal efficiencies of the cobalt and cesium by 0.01M of acetic acid were increased more than those by 0.05M of acetic acid due to the increase of the concrete zeta potential. The total effluent volume generated from the washing-electrokinetic decontamination was 11.55L (7.2ml/g).

Closure system of a vessel made of prestressed concrete

International Nuclear Information System (INIS)

Audibert, Alain

1974-01-01

The present invention relates to removable plugs of prestressed concrete which can be fitted to every type of closed high pressure vessels and especially to the cylindrical vessels of nuclear reactors. The method involved permits the plug to be fitted to the vessel through both radial and axial prestress. In this purpose, said invention proposes removable prestress ribs fitted inside sheaths in the plug and extending throughout the upper part of the bearing surfaces of the plug, said ribs being regularly arranged along the generators of an hyperboloid of one sheet. Owing to this important feature, that is to say said inclination of the ribs in accordance with the generators of said hyperboloid, said rib inclination can be changed on requirement for each realization [fr

The quest for performance-related specifications for hydraulic cement concrete.

Science.gov (United States)

1982-01-01

This paper reviews some of the problems associated with quality assurance for hydraulic cement concrete and the difficulties of relating the results of quality control and acceptance testing to the performance of the concrete facility. The importance...

West Valley Demonstration Project, Waste Management Area #3 -- Closure Alternative I

Energy Technology Data Exchange (ETDEWEB)

Marschke, Stephen F. [Environmental Measurements Laboratory (EML), New York, NY (United States)

2000-06-30

The Draft Environmental Impact Statement for the completion of the West Valley Demonstration Project and closure and/or long-term management of facilities at the Western New York Nuclear Service Center divided the site into Waste Management Areas (WMAs), and for each WMA, presented the impacts associated with five potential closure alternatives. This report focuses on WMA 3 (the High-Level Waste (HLW) Storage Area (Tanks 8D-1 and 8D-2), the Vitrification Facility and other facilities) and closure Alternative I (the complete removal of all structures, systems and components and the release of the area for unrestricted use), and reestimates the impacts associated with the complete removal of the HLW tanks, and surrounding facilities. A 32-step approach was developed for the complete removal of Tanks 8D-1 and 8D-2, the Supernatant Treatment System Support Building, and the Transfer Trench. First, a shielded Confinement Structure would be constructed to reduce the shine dose rate and to control radioactivity releases. Similarly, the tank heels would be stabilized to reduce potential radiation exposures. Next, the tank removal methodology would include: 1) excavation of the vault cover soil, 2) removal of the vault roof, 3) cutting off the tank’s top, 4) removal of the stabilized heel remaining inside the tank, 5) cutting up the tank’s walls and floor, 6) removal of the vault’s walls, the perlite blocks, and vault floor, and 7) radiation surveying and backfilling the resulting hole. After the tanks are removed, the Confinement Structure would be decontaminated and dismantled, and the site backfilled and landscaped. The impacts (including waste disposal quantities, emissions, work-effort, radiation exposures, injuries and fatalities, consumable materials used, and costs) were estimated based on this 32 step removal methodology, and added to the previously estimated impacts for closure of the other facilities within WMA 3 to obtain the total impacts from

Economics and risks of recycling radioactively contaminated concrete

International Nuclear Information System (INIS)

Parker, F.L.; Ayers, K.W.

1997-01-01

As Decontamination and Decommissioning activities proceed within the DOE complex, tremendous volumes of both radioactively contaminated and non-contaminated concrete will be processed for disposal. Current practice is to decontaminate the concrete, dispose of the contamination at LLW facilities and ship the concrete rubble to C ampersand D landfills for disposal. This study evaluates the economic, health and safety, legal, and social aspects of recycling radioactively contaminated concrete. Probabilistic models were used to estimate costs and risks. The model indicates that the radioactively contaminated concrete can be recycled at the same or lower cost than current or alternative practices. The risks associated with recycling were consistently less than or equal to the other alternatives considered

Concrete waste reduction of 50%

International Nuclear Information System (INIS)

Vos, R.M. de; Van der Wagt, K.M.; Van der Kruk, E.; Meeussen, H.W.

2016-01-01

During decommissioning quite a volume of concrete waste is produced. The degree of activation of the waste can range from clearly activated material to slightly activated or contaminated concrete. The degree of activation influences the applicable waste management processes that can be applied. The subsequent waste management processes can be identified for concrete waste are; disposal, segregation, re-use, conditional release and release. With each of these steps, the footprint of radioactive decommissioning waste is reduced. Future developments for concrete waste reduction can be achieved by applying smart materials in new build facilities (i.e. fast decaying materials). NRG (Nuclear Research and consultancy Group) has investigated distinctive waste management processes to reduce the foot-print of concrete waste streams resulting from decommissioning. We have investigated which processes can be applied in the Netherlands, both under current legislation and with small changes in legislation. We have also investigated the separation process in more detail. Pilot tests with a newly patented process have been started in 2015. We expect that our separation methods will reduce the footprint reduction of concrete waste by approximately 50% due to release or re-use in the nuclear sector or in the conventional industry. (authors)

Development of treatment technology for radioactive concrete wastes

Energy Technology Data Exchange (ETDEWEB)

Min, B. Y.; Choi, W. K.; Lee, K. W., E-mail: bymin@kaeri.re.k [Korea Atomic Energy Research Institute, 1045 Daeduk-daero, Yuseong-gu, Daejeon, 305-353 Republic of Korea (Korea, Republic of)

2010-10-15

The aim of this study was the separation of clean aggregates from contaminated dismantling concrete wastes by thermal and mechanical processes. In Korea, the decontamination and decommissioning of the retired Korea research reactor (KRR) and a uranium conversion plant (UCP) at the Korea Atomic Energy Research Institute (KAERI) has been under way. Hundreds of tons of concrete wastes are expected from these facilities. The KAERI has developed volume reduction technology applicable to an activated heavy concrete waste generated by dismantling KRR-2 and a uranium contaminated light weight concrete produced from a UCP. Contamination level of the gravel and sand aggregates was remarkably decreased by thermal and mechanical process. The volume reduction rate could be achieved above 70% for KRR-2 concrete waste and above to 80% for the UCP concrete waste. (Author)

Guidance for closure of existing DOE LLW disposal sites

International Nuclear Information System (INIS)

Blanchfield, L.

1987-01-01

During FY 1986, a closure guidance document was developed. The purpose of this document is to provide guidance in support of DOE Order 5820.2 to site operating contractors for the stabilization and closure of existing low-level waste (LLW) shallow land disposal sites at US Department of Energy (DOE) facilities. Guidance is provided to aid operators in placing existing LLW sites in a closed conditions, i.e., a condition in which a nonoperational site meets postclosure performance requirements and can be shown, within a high degree of confidence, to perform as anticipated in the future, under the most cost-effective maintenance approach. Guidance is based on the philosophy that closure should be planned and performed using a systems approach. Plans for FY 1987 call for revision of the document to incorporate more information on closure of LLW sites also containing radioactive mixed waste and/or transuranic waste. 4 references, 3 figures, 2 tables

Numerical simulation of a reinforced concrete shield around a nuclear reactor

International Nuclear Information System (INIS)

Mahama, Mumuni Salifu

1996-02-01

Ghana currently operates a Research Reactor and other nuclear facilities including a Gamma Irradiation Facility, a Radiographic Non-Destructive Testing laboratory and would be operating in the nearest future a Radiotherapy Centre. Each of these has a concrete radiation shield as a major safety device. In carrying out its functions, a concrete radiation shield may be subjected to thermal and mechanical stresses. A facility for analysing these stresses is desirable. Two computer codes have been developed under this programme for radiation shielding computation and stress analysis of cylindrical reactor shields. (au)

Multipass mining sequence room closures: In situ data report

International Nuclear Information System (INIS)

Munson, D.E.; Jones, R.L.; Northrop-Salazar, C.L.; Woerner, S.J.

1992-12-01

During the construction of the Thermal/Structural In Situ Test Rooms at the Waste Isolation Pilot Plant (WIPP) facility, measurements of the salt displacements were obtained at very early times, essentially concurrent with the mining activity. This was accomplished by emplacing manually read closure gage stations directly at the mining face, actually between the face and the mining machine, immediately upon mining of the intended gage location. Typically, these mining sequence closure measurements were taken within one hour of mining of the location and within one meter of the mining face. Readings were taken at these gage stations as the multipass mining continued, with the gage station reestablished as each successive mining pass destroyed the earlier gage points. Data reduction yields the displacement history during the mining operation. These early mining sequence closure data, when combined with the later data of the permanently emplaced closure gages, gives the total time-dependent closure displacements of the test rooms. This complete closure history is an essential part of assuring that the in situ test databases will provide an adequate basis for validation of the predictive technology of salt creep behavior, as required by the WIPP technology development program for disposal of radioactive waste in bedded salt

Radionuclide Retention in Concrete Waste Forms

Energy Technology Data Exchange (ETDEWEB)

Mattigod, Shas V.; Bovaird, Chase C.; Wellman, Dawn M.; Wood, Marcus I.

2010-09-30

Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation); the mechanism of contaminant release; the significance of contaminant release pathways; how waste form performance is affected by the full range of environmental conditions within the disposal facility; the process of waste form aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of waste form aging on chemical, physical, and radiological properties; and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the waste forms come in contact with groundwater. The information presented in the report provides data that 1) quantify radionuclide retention within concrete waste form materials similar to those used to encapsulate waste in the Low-Level Waste Burial Grounds (LLBG); 2) measure the effect of concrete waste form properties likely to influence radionuclide migration; and 3) quantify the stability of uranium-bearing solid phases of limited solubility in concrete.

Single-shell tank closure work plan. Revision A

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-06-01

In January 1994, the Hanford Federal Facility Agreement and Conset Order (Tri-Party Agreement) was amended to reflect a revised strategy for remediation of radioactive waste in underground storage tanks. These amendments include milestones for closure of the single-shell tank (SST) operable units, to be initiated by March 2012 and completed by September 2024. This SST-CWP has been prepared to address the principal topical areas identified in Tri-Party Agreement Milestone M-45-06 (i.e., regulatory pathway, operable unit characterization, waste retrieval, technology development, and a strategy for achieving closure). Chapter 2.0 of this SST-CWP provides a brief description of the environmental setting, SST System, the origin and characteristics of SST waste, and ancillary equipment that will be remediated as part of SST operable unit closure. Appendix 2A provides a description of the hydrogeology of the Hanford Site, including information on the unsaturated sediments (vadose zone) beneath the 200 Areas Plateau. Chapter 3.0 provides a discussion of the laws and regulations applicable to closure of the SST farm operable units. Chapter 4.0 provides a summary description of the ongoing characterization activities that best align with the proposed regulatory pathway for closure. Chapter 5.0 describes aspects of the SST waste retrieval program, including retrieval strategy, technology, and sequence, potential tank leakage during retrieval, and considerations of deployment of subsurface barriers. Chapter 6.0 outlines a proposed strategy for closure. Chapter 7.0 provides a summary of the programs underway or planned to develop technologies to support closure. Ca. 325 refs.

Single-shell tank closure work plan. Revision A

International Nuclear Information System (INIS)

1995-06-01

In January 1994, the Hanford Federal Facility Agreement and Conset Order (Tri-Party Agreement) was amended to reflect a revised strategy for remediation of radioactive waste in underground storage tanks. These amendments include milestones for closure of the single-shell tank (SST) operable units, to be initiated by March 2012 and completed by September 2024. This SST-CWP has been prepared to address the principal topical areas identified in Tri-Party Agreement Milestone M-45-06 (i.e., regulatory pathway, operable unit characterization, waste retrieval, technology development, and a strategy for achieving closure). Chapter 2.0 of this SST-CWP provides a brief description of the environmental setting, SST System, the origin and characteristics of SST waste, and ancillary equipment that will be remediated as part of SST operable unit closure. Appendix 2A provides a description of the hydrogeology of the Hanford Site, including information on the unsaturated sediments (vadose zone) beneath the 200 Areas Plateau. Chapter 3.0 provides a discussion of the laws and regulations applicable to closure of the SST farm operable units. Chapter 4.0 provides a summary description of the ongoing characterization activities that best align with the proposed regulatory pathway for closure. Chapter 5.0 describes aspects of the SST waste retrieval program, including retrieval strategy, technology, and sequence, potential tank leakage during retrieval, and considerations of deployment of subsurface barriers. Chapter 6.0 outlines a proposed strategy for closure. Chapter 7.0 provides a summary of the programs underway or planned to develop technologies to support closure. Ca. 325 refs

Tank Closure Progress at the Department of Energy's Idaho National Engineering Laboratory Tank Farm Facility

International Nuclear Information System (INIS)

Butterworth, St.W.; Shaw, M.R.

2009-01-01

Significant progress continued at the U.S. Department of Energy (DOE) Idaho National Laboratory (INL) with the completion of the closure process to empty, clean and close radioactive liquid waste storage tanks at the Idaho Nuclear Technology and Engineering Center (INTEC) Tank Farm Facility (TFF). The TFF includes eleven 1,135.6-kL (300,000-gal) underground stainless steel storage tanks and four smaller, 113.5-kL (30,000-gal) stainless steel tanks, along with tank vaults, interconnecting piping, and ancillary equipment. The TFF tanks had historically been used to store a variety of radioactive liquid waste, including wastes associated with past spent nuclear fuel reprocessing. Four of the large storage tanks remain in use for waste storage while the other seven 1,135.6-kL (300,000-gal) tanks and the four 113.5-kL (30,000-gal) tanks have been emptied of waste, cleaned and filled with grout. Recent issuance of an Amended Record of Decision (ROD) in accordance with the National Environmental Policy Act, and a Waste Determination complying with Section 3116 of the Ronald W. Reagan National Defense Authorization Act (NDAA) for Fiscal Year 2005, allowed commencement of grouting activities on the cleaned tanks. The first three 113.5-kL (30,000-gal) tanks were grouted in the Fall of 2006 and the fourth tank and the seven 1,135.6-kL (300,000-gal) tanks were filled with grout in 2007 to provide long-term stability. During 2008 over seven miles of underground process piping along with associated tank valve boxes and secondary containment systems was stabilized with grout. Lessons learned were compiled and implemented during the closure process and will be utilized on the remaining four 1,135.6-kL (300,000-gal) underground stainless steel storage tanks. Significant progress has been made to clean and close emptied tanks at the INTEC TFF. Between 2002 and 2005, seven of the eleven 1,135.6-kL (300,000-gal) tanks and all four 113.5-kL (30,000-gal) tanks were cleaned and prepared

40 CFR 265.118 - Post-closure plan; amendment of plan.

Science.gov (United States)

2010-07-01

... mail. In addition, for facilities without approved post-closure plans, it must also be provided during... requirements. At the end of the specified period of suspension, the Regional Ad-min-is-tra-tor would then...

The impact of rural hospital closures on equity of commuting time for haemodialysis patients: simulation analysis using the capacity-distance model.

Science.gov (United States)

Matsumoto, Masatoshi; Ogawa, Takahiko; Kashima, Saori; Takeuchi, Keisuke

2012-07-23

Frequent and long-term commuting is a requirement for dialysis patients. Accessibility thus affects their quality of lives. In this paper, a new model for accessibility measurement is proposed in which both geographic distance and facility capacity are taken into account. Simulation of closure of rural facilities and that of capacity transfer between urban and rural facilities are conducted to evaluate the impacts of these phenomena on equity of accessibility among dialysis patients. Post code information as of August 2011 of all the 7,374 patients certified by municipalities of Hiroshima prefecture as having first or third grade renal disability were collected. Information on post code and the maximum number of outpatients (capacity) of all the 98 dialysis facilities were also collected. Using geographic information systems, patient commuting times were calculated in two models: one that takes into account road distance (distance model), and the other that takes into account both the road distance and facility capacity (capacity-distance model). Simulations of closures of rural and urban facilities were then conducted. The median commuting time among rural patients was more than twice as long as that among urban patients (15 versus 7 minutes, psimulation, when five rural public facilitiess were closed, Gini coefficient of commuting times among the patients increased by 16%, indicating a substantial worsening of equity, and the number of patients with commuting times longer than 90 minutes increased by 72 times. In contrast, closure of four urban public facilities with similar capacities did not affect these values. Closures of dialysis facilities in rural areas have a substantially larger impact on equity of commuting times among dialysis patients than closures of urban facilities. The accessibility simulations using the capacity-distance model will provide an analytic framework upon which rational resource distribution policies might be planned.

Evaluation of the concrete shield compositions from the 2010 criticality accident alarm system benchmark experiments at the CEA Valduc SILENE facility

International Nuclear Information System (INIS)

Miller, Thomas Martin; Celik, Cihangir; Dunn, Michael E; Wagner, John C; McMahan, Kimberly L; Authier, Nicolas; Jacquet, Xavier; Rousseau, Guillaume; Wolff, Herve; Savanier, Laurence; Baclet, Nathalie; Lee, Yi-kang; Trama, Jean-Christophe; Masse, Veronique; Gagnier, Emmanuel; Naury, Sylvie; Blanc-Tranchant, Patrick; Hunter, Richard; Kim, Soon; Dulik, George Michael; Reynolds, Kevin H.

2015-01-01

In October 2010, a series of benchmark experiments were conducted at the French Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the United States Department of Energy Nuclear Criticality Safety Program and the CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems. This series of experiments consisted of three single-pulsed experiments with the SILENE reactor. For the first experiment, the reactor was bare (unshielded), whereas in the second and third experiments, it was shielded by lead and polyethylene, respectively. The polyethylene shield of the third experiment had a cadmium liner on its internal and external surfaces, which vertically was located near the fuel region of SILENE. During each experiment, several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor. Nearly half of the foils and TLDs had additional high-density magnetite concrete, high-density barite concrete, standard concrete, and/or BoroBond shields. CEA Saclay provided all the concrete, and the US Y-12 National Security Complex provided the BoroBond. Measurement data from the experiments were published at the 2011 International Conference on Nuclear Criticality (ICNC 2011) and the 2013 Nuclear Criticality Safety Division (NCSD 2013) topical meeting. Preliminary computational results for the first experiment were presented in the ICNC 2011 paper, which showed poor agreement between the computational results and the measured values of the foils shielded by concrete. Recently the hydrogen content, boron content, and density of these concrete shields were further investigated within the constraints of the previously available data. New computational results for the first experiment are now available

Surface Wave Velocity-Stress Relationship in Uniaxially Loaded Concrete

DEFF Research Database (Denmark)

Shokouhi, Parisa; Zoëga, Andreas; Wiggenhauser, Herbert

2012-01-01

The sonic surface wave (or Rayleigh wave) velocity measured on prismatic concrete specimens under uniaxial compression was found to be highly stress-dependent. At low stress levels, the acoustoelastic effect and the closure of existing microcracks results in a gradual increase in surface wave...... velocities. At higher stress levels, concrete suffers irrecoverable damage: the existing microcracks widen and coalesce and new microcracks form. This progressive damage process leads first to the flattening and eventually the drop in the velocity-stress curves. Measurements on specimens undergoing several...... loading cycles revealed that the velocities show a stress-memory effect in good agreement with the Kaiser effect. Comparing the velocities measured during loading and unloading, the effects of stress and damage on the measured velocities could be differentiated. Moreover, the stress dependency of surface...

A study on closure performance in geological disposal of high-level radioactive waste (H14)

International Nuclear Information System (INIS)

Sugita, Yutaka; Kawakami, Susumu; Yui, Mikazu; Makino, Hitoshi; Sawada, Atsushi; Kurihara, Yuji; Mihara, Morihiro

2003-04-01

Regarding closure technology of underground facilities in geological disposal of the HLW in H12 report, the fundamental concept that closure technology has no impact against the engineered barrier system (EBS) was described. Performance Assessment (PA) has been performed without considering of the barrier function of closure elements. Following H12 report, the various in-situ data of the closure elements (ex. plug, backfill) have been obtained. Therefore, we considered that the PA of the EBS considering the expecting performance of the closure elements from the view points of both the engineering technology and the PA should be examined. First, the characteristics of rock mass and the function of the closure elements were summarized. Then, the closure scenario was developed preliminarily based on hydrological analysis between a hydraulic fracture and a disposal panel, the fault tree analysis, and so on. (author)

Special feature of the facilities for final disposal of radioactive waste and its potential impact on the licensing process

International Nuclear Information System (INIS)

Lee Gonzales, Horacio M.; Medici, Marcela A.; Alvarez, Daniela E.; Biaggio, Alfredo L.

2009-01-01

During the lifetime of a radioactive waste disposal facility it is possible to identify five stages: design, construction, operation, closure and post-closure. While the design, and pre-operation stages are, to some extent, similar to other kind of nuclear or radioactive facilities; construction, operation, closure and post-closure have quite special meanings in the case of radioactive waste disposal systems. For instance, the 'closure' stage of a final disposal facility seems to be equivalent to the commissioning stage of a conventional nuclear or radioactive facility. This paper describes the unique characteristics of these stages of final disposal systems, that lead to concluded that their licensing procedure can not be assimilated to the standard licensing procedures in use for other nuclear or radioactive facilities, making it necessary to develop a tailored license system. (author)

Closure End States for Facilities, Waste Sites, and Subsurface Contamination - 12543

Energy Technology Data Exchange (ETDEWEB)

Gerdes, Kurt; Chamberlain, Grover; Whitehurst, Latrincy; Marble, Justin [Office of Groundwater and Soil Remediation, U.S. Department of Energy, Washington, DC 20585 (United States); Wellman, Dawn [Pacific Northwest National Laboratory, Richland, Washington 99352 (United States); Deeb, Rula; Hawley, Elisabeth [ARCADIS U.S., Inc., Emeryville, CA 94608 (United States)

2012-07-01

The United States (U.S.) Department of Energy (DOE) manages the largest groundwater and soil cleanup effort in the world. DOE's Office of Environmental Management (EM) has made significant progress in its restoration efforts at sites such as Fernald and Rocky Flats. However, remaining sites, such as Savannah River Site, Oak Ridge Site, Hanford Site, Los Alamos, Paducah Gaseous Diffusion Plant, Portsmouth Gaseous Diffusion Plant, and West Valley Demonstration Project possess the most complex challenges ever encountered by the technical community and represent a challenge that will face DOE for the next decade. Closure of the remaining 18 sites in the DOE EM Program requires remediation of 75 million cubic yards of contaminated soil and 1.7 trillion gallons of contaminated groundwater, deactivation and decommissioning (D and D) of over 3000 contaminated facilities and thousands of miles of contaminated piping, removal and disposition of millions of cubic yards of legacy materials, treatment of millions of gallons of high level tank waste and disposition of hundreds of contaminated tanks. The financial obligation required to remediate this volume of contaminated environment is estimated to cost more than 7% of the to-go life-cycle cost. Critical in meeting this goal within the current life-cycle cost projections is defining technically achievable end states that formally acknowledge that remedial goals will not be achieved for a long time and that residual contamination will be managed in the interim in ways that are protective of human health and environment. Formally acknowledging the long timeframe needed for remediation can be a basis for establishing common expectations for remedy performance, thereby minimizing the risk of re-evaluating the selected remedy at a later time. Once the expectations for long-term management are in place, remedial efforts can be directed towards near-term objectives (e.g., reducing the risk of exposure to residual contamination

Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 116: Area 25 Test Cell C Facility, Nevada Test Site, Nevada, Revision 1

International Nuclear Information System (INIS)

2008-01-01

This Streamlined Approach for Environmental Restoration (SAFER) Plan identifies the activities required for the closure of Corrective Action Unit (CAU) 116, Area 25 Test Cell C Facility. The Test Cell C (TCC) Facility is located in Area 25 of the Nevada Test Site (NTS) approximately 25 miles northwest of Mercury, Nevada (Figure 1). CAU 116 is currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO) of 1996 (as amended February 2008) and consists of two Corrective Action Sites (CASs): (1) CAS 25-23-20, Nuclear Furnace Piping; and (2) CAS 25-41-05, Test Cell C Facility. CAS 25-41-05 is described in the FFACO as the TCC Facility but actually includes Building 3210 and attached concrete shield wall only. CAU 116 will be closed by demolishing Building 3210, the attached concrete shield wall, and the nuclear furnace piping. In addition, as a best management practice (BMP), Building 3211 (moveable shed) will be demolished due to its close proximity to Building 3210. This will aid in demolition and disposal operations. Radiological surveys will be performed on the demolition debris to determine the proper disposal pathway. As much of the demolition debris as space allows will be placed into the Building 3210 basement structure. After filling to capacity with demolition debris, the basement structure will be mounded or capped and closed with administrative controls. Prior to beginning demolition activities and according to an approved Sampling and Analysis Plan (SAP), representative sampling of surface areas that are known, suspected, or have the potential to contain hazardous constituents such as lead or polychlorinated biphenyls (PCBs) will be performed throughout all buildings and structures. Sections 2.3.2, 4.2.2.2, 4.2.2.3, 4.3, and 6.2.6.1 address the methodologies employed that assure the solid debris placed in the basement structure will not contain contaminants of concern (COCs) above hazardous waste levels. The anticipated post-closure

Effect of Concrete Waste Form Properties on Radionuclide Migration

International Nuclear Information System (INIS)

Mattigod, Shas V.; Bovaird, Chase C.; Wellman, Dawn M.; Skinner, De'Chauna J.; Cordova, Elsa A.; Wood, Marcus I.

2009-01-01

Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation) the mechanism of contaminant release, the significance of contaminant release pathways, how waste form performance is affected by the full range of environmental conditions within the disposal facility, the process of waste form aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility, the effect of waste form aging on chemical, physical, and radiological properties and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the waste forms come in contact with groundwater. Numerous sets of tests were initiated in fiscal years (FY) 2006-2009 to evaluate (1) diffusion of iodine (I) and technetium (Tc) from concrete into uncontaminated soil after 1 and 2 years, (2) I and rhenium (Re) diffusion from contaminated soil into fractured concrete, (3) I and Re (set 1) and Tc (set 2) diffusion from fractured concrete into uncontaminated soil, (4) evaluate the moisture distribution profile within the sediment half-cell, (5) the reactivity and speciation of uranium (VI) (U(VI)) compounds in concrete porewaters, (6) the rate of dissolution of concrete monoliths, and (7) the diffusion of simulated tank waste into concrete.

Closure Report for Corrective Action Unit 425: Area 9 Main Lake Construction Debris Disposal Area, Tonopah Test Range, Nevada

Energy Technology Data Exchange (ETDEWEB)

K. B. Campbell

2003-03-01

Corrective Action Unit (CAU) 425 is located on the Tonopah Test Range, approximately 386 kilometers (240 miles) northwest of Las Vegas, Nevada. CAU 425 is listed in the Federal Facility Agreement and Consent Order (FFACO, 1996) and is comprised of one Corrective Action Site (CAS). CAS 09-08-001-TA09 consisted of a large pile of concrete rubble from the original Hard Target and construction debris associated with the Tornado Rocket Sled Tests. CAU 425 was closed in accordance with the FFACO and the Nevada Division of Environmental Protection-approved Streamlined Approach for Environmental Restoration Plan for CAU 425: Area 9 Main Lake Construction Debris Disposal Area, Tonopah Test Range, Nevada (U.S. Department of Energy, Nevada Operations Office, 2002). CAU 425 was closed by implementing the following corrective actions: The approved corrective action for this unit was clean closure. Closure activities included: (1) Removal of all the debris from the site. (2) Weighing each load of debris leaving the job site. (3) Transporting the debris to the U.S. Air Force Construction Landfill for disposal. (4) Placing the radioactive material in a U.S. Department of Transportation approved container for proper transport and disposal. (5) Transporting the radioactive material to the Nevada Test Site for disposal. (6) Regrading the job site to its approximate original contours/elevation.

Modelling of multiphase flow in concrete cells of the radioactive waste storage facility at El Cabril (Spain)

International Nuclear Information System (INIS)

Chaparro, M.C.; Saaltink, M.W.

2015-01-01

El Cabril is the low and intermediate level radioactive waste disposal facility for Spain. After sealing the cells that stored the radioactive waste, water was collected from a drainpipe, indicating the flow of water within the cell. A hypothesis had been proposed to explain this phenomenon which consists of capillary rise from groundwater and evaporation and condensation within the cell produced by temperature gradients caused by seasonal temperature fluctuations outside. To corroborate this hypothesis a 2D numerical model was made taking into account all relevant processes such as multiphase flow and heat transport. Data were used measured by sensors in the cells and data from laboratory test. There is a good agreement between the temperature measured by the sensors and the ones calculated by the model. The model shows a drying of the concrete at the hot side (that is the wall during summer and the container during winter). The concrete is saturated with water at the cold side (that is the container during summer and the wall in winter), leading to runoff of water to the drainpipe. The flux at this drainpipe occurred in the two yearly periods, being higher in winter than in summer. (authors)

Tumulus Disposal Demonstration Facility for the Oak Ridge Reservation

International Nuclear Information System (INIS)

Clapp, R.B.; van Hoesen, S.D.

1987-01-01

This disposal concept is based on the Tumulus design developed by the French at the La Manche facility. Waste units are stacked above-grade on a concrete pad. The facility currently under development at the Oak Ridge National Laboratory (ORNL) involves sealing waste in concrete vaults, placing the vaults on a grade level concrete pad, and covering the pad and vaults with a soil cover after vault emplacement is complete. Emplacement is expected to continue until the facility exhausts its approximate 800 m 3 (28,000 ft 3 ) capacity. The facility incorporates engineered barriers to radionuclide migration; a monitoring system to ensure barrier performance; and a newly developed set of Demonstration Waste Acceptance Criteria to reduce the likelihood of groundwater contamination

Concrete shielding for nuclear ship 'Mutsu'

International Nuclear Information System (INIS)

Nagase, Tetsuo; Nakajima, Tadao; Okumura, Tadahiko; Saito, Tetsuo

1983-01-01

The nuclear ship ''Mutsu'' was constructed in 1970 as the fourth in the world. On September 1, 1974, during the power raising test in the Pacific Ocean, radiation leak was detected. As the result of investigation, it was found that the cause was the fast neutrons streaming through the gap between the reactor pressure vessel and the primary shield. In order to repair the shielding facility, the Japan Nuclear Ship Research Development Agency carried out research and development and shielding design. It was decided to adopt serpentine concrete for the primary shield, which is the excellent moderator of fast neutrons even at high temperature, and heavy concrete for the secondary shield, which is effective for shielding both gamma ray and neutron beam. The repair of shielding was carried out in the Sasebo Shipyard, and completed in August, 1982. The outline of the repair work is reported. The weight increase was about 300 t. The conditions of the shielding design, the method of shielding analysis, the performance required for the shielding concrete, the preliminary experiment on heavy concrete and the construction works of serpentine concrete and heavy concrete are described. (Kako, I.)

Preliminary engineering and economic analysis of the fixation of high-level radioactive wastes in concrete

International Nuclear Information System (INIS)

Weeren, H.O.; Perona, J.J.

1979-07-01

This feasibility study was based on a waste fixation facility that would serve a reprocessing plant with a capacity of 5 metric tons of uranium per day (MTU/day). Postirradiation cooling times of 3 to 10 years prior to waste solidification were assumed. The waste solution would be concentrated, denitrated, mixed with cement, and cast under pressure in cylindrical canisters similar to those envisioned for a glass facility. The solidified waste grout would be vented, to allow the free water to escape, and then sealed. The filled canisters would be shipped to a geologic repository for permanent storage. Recent work with concretes formed under elevated temperatures and pressures (FUETAP) indicates that they are highly leach resistant. The operating costs were estimated for a waste fixation facility under several conditions. Operating costs for a glass fixation facility were also estimated and compared with the operating costs for a concrete fixation facility. The principal conclusion is that concrete could be an alternative to glass as a matrix for fixation of wastes with high heat-generation rates. The operating costs of an optimized concrete fixation process would probably not be greatly higher than the operating costs of a glass plant, and the capital costs would almost surely be lower. In addition, the concrete process is not a high-temperature process and would not have the consequent operating problems

Diffusion and Leaching Behavior of Radionuclides in Category 3 Waste Encasement Concrete and Soil Fill Material – Summary Report

Energy Technology Data Exchange (ETDEWEB)

Mattigod, Shas V.; Wellman, Dawn M.; Bovaird, Chase C.; Parker, Kent E.; Clayton, Libby N.; Powers, Laura; Recknagle, Kurtis P.; Wood, Marcus I.

2011-08-31

One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Such concrete encasement would contain and isolate the waste packages from the hydrologic environment and would act as an intrusion barrier. The current plan for waste isolation consists of stacking low-level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages in concrete. These concrete-encased waste stacks are expected to vary in size with maximum dimensions of 6.4 m long, 2.7 m wide, and 4 m high. The waste stacks are expected to have a surrounding minimum thickness of 15 cm of concrete encasement. These concrete-encased waste packages are expected to withstand environmental exposure (solar radiation, temperature variations, and precipitation) until an interim soil cover or permanent closure cover is installed, and to remain largely intact thereafter. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. The mobilized radionuclides may escape from the encased concrete by mass flow and/or diffusion and move into the surrounding subsurface environment. Therefore, it is necessary to assess the performance of the concrete encasement structure and the ability of the surrounding soil to retard radionuclide migration. The retardation factors for radionuclides contained in the waste packages can be determined from measurements of diffusion coefficients for these contaminants through concrete and fill material. Some of the mobilization scenarios include (1) potential leaching of waste form before permanent closure cover is installed; (2) after the cover installation, long-term diffusion of radionuclides from concrete waste form into surrounding fill material; (3) diffusion of radionuclides from contaminated soils into adjoining concrete encasement and clean fill material. Additionally, the rate of

Implementing RCRA during facility deactivation

International Nuclear Information System (INIS)

Lebaron, G.J.

1997-01-01

RCRA regulations require closure of permitted treatment, storage and disposal (TSD) facilities within 180 days after cessation of operations, and this may essentially necessitate decommissioning to complete closure. A more cost effective way to handle the facility would be to significantly reduce the risk to human health and the environment by taking it from its operational status to a passive, safe, inexpensive-to-maintain surveillance and maintenance condition (deactivation) prior to decommissioning. This paper presents an innovative approach to the cost effective deactivation of a large, complex chemical processing facility permitted under RCRA. The approach takes into account risks to the environment posed by this facility in comparison to risks posed by neighboring facilities at the site. The paper addresses the manner in which: 1) stakeholders and regulators were involved; 2) identifies a process by which the project proceeds and regulators and stakeholders were involved; 3) end points were developed so completion of deactivation was clearly identified at the beginning of the project, and 4) innovative practices were used to deactivate more quickly and cost effectively

Radionuclide Retention in Concrete Wasteforms

Energy Technology Data Exchange (ETDEWEB)

Bovaird, Chase C.; Jansik, Danielle P.; Wellman, Dawn M.; Wood, Marcus I.

2011-09-30

Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation); the mechanism of contaminant release; the significance of contaminant release pathways; how wasteform performance is affected by the full range of environmental conditions within the disposal facility; the process of wasteform aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of wasteform aging on chemical, physical, and radiological properties; and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the wasteforms come in contact with groundwater. The information present in the report provides data that (1) measures the effect of concrete wasteform properties likely to influence radionuclide migration; and (2) quantifies the rate of carbonation of concrete materials in a simulated vadose zone repository.

105-DR Large Sodium Fire Facility decontamination, sampling, and analysis plan

International Nuclear Information System (INIS)

Knaus, Z.C.

1995-01-01

This is the decontamination, sampling, and analysis plan for the closure activities at the 105-DR Large Sodium Fire Facility at Hanford Reservation. This document supports the 105-DR Large Sodium Fire Facility Closure Plan, DOE-RL-90-25. The 105-DR LSFF, which operated from about 1972 to 1986, was a research laboratory that occupied the former ventilation supply room on the southwest side of the 105-DR Reactor facility in the 100-D Area of the Hanford Site. The LSFF was established to investigate fire fighting and safety associated with alkali metal fires in the liquid metal fast breeder reactor facilities. The decontamination, sampling, and analysis plan identifies the decontamination procedures, sampling locations, any special handling requirements, quality control samples, required chemical analysis, and data validation needed to meet the requirements of the 105-DR Large Sodium Fire Facility Closure Plan in compliance with the Resource Conservation and Recovery Act

The impact of rural hospital closures on equity of commuting time for haemodialysis patients: simulation analysis using the capacity-distance model

Directory of Open Access Journals (Sweden)

Matsumoto Masatoshi

2012-07-01

Full Text Available Abstract Background Frequent and long-term commuting is a requirement for dialysis patients. Accessibility thus affects their quality of lives. In this paper, a new model for accessibility measurement is proposed in which both geographic distance and facility capacity are taken into account. Simulation of closure of rural facilities and that of capacity transfer between urban and rural facilities are conducted to evaluate the impacts of these phenomena on equity of accessibility among dialysis patients. Methods Post code information as of August 2011 of all the 7,374 patients certified by municipalities of Hiroshima prefecture as having first or third grade renal disability were collected. Information on post code and the maximum number of outpatients (capacity of all the 98 dialysis facilities were also collected. Using geographic information systems, patient commuting times were calculated in two models: one that takes into account road distance (distance model, and the other that takes into account both the road distance and facility capacity (capacity-distance model. Simulations of closures of rural and urban facilities were then conducted. Results The median commuting time among rural patients was more than twice as long as that among urban patients (15 versus 7 minutes, p  Conclusions Closures of dialysis facilities in rural areas have a substantially larger impact on equity of commuting times among dialysis patients than closures of urban facilities. The accessibility simulations using thecapacity-distance model will provide an analytic framework upon which rational resource distribution policies might be planned.

Closure Report for Corrective Action Unit 543: Liquid Disposal Units, Nevada Test Site, Nevada

Energy Technology Data Exchange (ETDEWEB)

NSTec Environmental Restoration

2008-01-01

This Closure Report (CR) documents closure activities for Corrective Action Unit (CAU) 543, Liquid Disposal Units, according to the Federal Facility Agreement and Consent Order (FFACO, 1996) and the Corrective Action Plan (CAP) for CAU 543 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2007). CAU 543 is located at the Nevada Test Site (NTS), Nevada (Figure 1), and consists of the following seven Corrective Action Sites (CASs): CAS 06-07-01, Decon Pad; CAS 15-01-03, Aboveground Storage Tank; CAS 15-04-01, Septic Tank; CAS 15-05-01, Leachfield; CAS 15-08-01, Liquid Manure Tank; CAS 15-23-01, Underground Radioactive Material Area; CAS 15-23-03, Contaminated Sump, Piping; and CAS 06-07-01 is located at the Decontamination Facility in Area 6, adjacent to Yucca Lake. The remaining CASs are located at the former U.S. Environmental Protection Agency (EPA) Farm in Area 15. The purpose of this CR is to provide a summary of the completed closure activities, to document waste disposal, and to present analytical data confirming that the remediation goals were met. The closure alternatives consisted of closure in place for two of the CASs, and no further action with implementation of best management practices (BMPs) for the remaining five CASs.

Research reactor decommissioning experience - concrete removal and disposal -

International Nuclear Information System (INIS)

Manning, Mark R.; Gardner, Frederick W.

1990-01-01

Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limits for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations

E AREA LOW LEVEL WASTE FACILITY DOE 435.1 PERFORMANCE ASSESSMENT

Energy Technology Data Exchange (ETDEWEB)

Wilhite, E

2008-03-31

This Performance Assessment for the Savannah River Site E-Area Low-Level Waste Facility was prepared to meet requirements of Chapter IV of the Department of Energy Order 435.1-1. The Order specifies that a Performance Assessment should provide reasonable assurance that a low-level waste disposal facility will comply with the performance objectives of the Order. The Order also requires assessments of impacts to water resources and to hypothetical inadvertent intruders for purposes of establishing limits on radionuclides that may be disposed near-surface. According to the Order, calculations of potential doses and releases from the facility should address a 1,000-year period after facility closure. The point of compliance for the performance measures relevant to the all pathways and air pathway performance objective, as well as to the impact on water resources assessment requirement, must correspond to the point of highest projected dose or concentration beyond a 100-m buffer zone surrounding the disposed waste following the assumed end of active institutional controls 100 years after facility closure. During the operational and institutional control periods, the point of compliance for the all pathways and air pathway performance measures is the SRS boundary. However, for the water resources impact assessment, the point of compliance remains the point of highest projected dose or concentration beyond a 100-m buffer zone surrounding the disposed waste during the operational and institutional control periods. For performance measures relevant to radon and inadvertent intruders, the points of compliance are the disposal facility surface for all time periods and the disposal facility after the assumed loss of active institutional controls 100 years after facility closure, respectively. The E-Area Low-Level Waste Facility is located in the central region of the SRS known as the General Separations Area. It is an elbow-shaped, cleared area, which curves to the northwest

Historic Concrete : From Concrete Repair to Concrete Conservation

NARCIS (Netherlands)

Heinemann, H.A.

2013-01-01

Concrete like materials were already applied during the Roman Empire. After the decline of the Roman Empire, a wide scale application of concrete only reappeared in the 19th century. Here lies also the origin of modern (reinforced) concrete. Since then, both concrete application and composition have

Cutting techniques of reinforced concrete by wire sawing

International Nuclear Information System (INIS)

Miyao, Hidehiko; Komatsu, Junji; Kamiyama, Yoshinori; Yasoshima, Harunori; Kukino, Yoshinori; Yamamoto, Yuichi; Miyazaki, Takashi; Aritomi, Masanori

1995-01-01

The Research Association for Nuclear Facility Decommissioning (RANDEC) has been carrying out demonstration tests to improve current technologies for decommissioning. The conceptual dismantling system has been studied and basic cutting tests have been carried out by wire sawing. In terms of waste management and dismantling efficiency, the diamond wire saw cutting method has advantages for cutting radioactive concrete in large blocks. A conceptual design for a dismantling system for various concrete shieldings of nuclear facilities has been developed and diamond wire sawing has been designed and manufactured. The basic cutting tests by wire sawing have been carried out to obtain quantitative data, in addition to the conceptual design of a dismantling system for biological shielding of various power reactors (PWR, BWR, GCR) and cell walls of nuclear fuel cycle facilities. On the basis of the conceptual dismantling system and quantitative cutting performance data, wire sawing equipment has been manufactured for use in nuclear facilities. This study was performed on consignment for the Science and Technology Agency of Japan. (author)

Cylindrical prestressed concrete pressure vessel for a nuclear power plant

International Nuclear Information System (INIS)

Horner, M.; Hodzic, A.; Haferkamp, D.

1976-01-01

A prestressed concrete pressure vessel for a HTGR is proposed which encloses, in addition to the reactor core, not only the heat-exchanging facilities but also the turbine unit. The reinforcement of the cylindrical concrete body is to be carried out with special care, it is provided for horizontal tendons, the prestressed concrete pressure vessel has a wire-winding device, while the longitudinal reinforcement is achieved by tendous guided in parallel to the vesses axes through the interspaces between the pods. (UWI) [de

Derivation of activity limits for the disposal of radioactive waste in near surface disposal facilities

International Nuclear Information System (INIS)

2003-12-01

Radioactive waste must be managed safely, consistent with internationally agreed safety standards. The disposal method chosen for the waste should be commensurate with the hazard and longevity of the waste. Near surface disposal is an option used by many countries for the disposal of radioactive waste containing mainly short lived radionuclides and low concentrations of long lived radionuclides. The term 'near surface disposal' encompasses a wide range of design options, including disposal in engineered structures at or just below ground level, disposal in simple earthen trenches a few metres deep, disposal in engineered concrete vaults, and disposal in rock caverns several tens of metres below the surface. The use of a near surface disposal option requires design and operational measures to provide for the protection of human health and the environment, both during operation of the disposal facility and following its closure. To ensure the safety of both workers and the public (both in the short term and the long term), the operator is required to design a comprehensive waste management system for the safe operation and closure of a near surface disposal facility. Part of such a system is to establish criteria for accepting waste for disposal at the facility. The purpose of the criteria is to limit the consequences of events which could lead to radiation exposures and in addition, to prevent or limit hazards, which could arise from non-radiological causes. Waste acceptance criteria include limits on radionuclide content concentration in waste materials, and radionuclide amounts in packages and in the repository as a whole. They also include limits on quantity of free liquids, requirements for exclusion of chelating agents and pyrophoric materials, and specifications of the characteristics of the waste containers. Largely as a result of problems encountered at some disposal facilities operated in the past, in 1985 the IAEA published guidance on generic acceptance

Safety at the End of a Nuclear Facility's Life

International Nuclear Information System (INIS)

Geis, John A.; McEahern, Patrice; Evans, Brad

2004-01-01

The objective of this paper is to capture the changes that are caused by the transition from nuclear operation through closure of defense nuclear facilities and convey lessons learned from their deactivation, decontamination and demolition. The specific area of discussion is focused on the planned reduction of safety equipment and consequent shift in hazard controls and safety management programs as the facility moves toward closure. The premise of the paper is that as the dominant hazards transition from nuclear to radiological and/or industrial, the facility control of the hazards and response to the potential upset conditions must transition as well to ensure safe and efficient operations. Using recent experience of the accelerated closure mission for U. S. Department of Energy (DOE) defense nuclear facilities at Rocky Flats Environmental Technology Site, the current culture with respect to developing and implementing hazard controls and response to upset conditions is illustrated. Several events have been documented that provide insight into the challenges facing line managers and safety professionals at the end of a facility's life cycle. Replacing permanent systems with temporary equipment challenges the traditional concept of reliability. Workers disassemble safety systems daily, but must rely on some of these components or redundant systems as work continues. Decisions governing upkeep of systems that await demolition balance the risk of running to failure against the cost benefit of maintenance and repair. This is further complicated as regulators and safety professionals are often unfamiliar with these new conditions and continue to view facility work activities and potential upset conditions from a nuclear operations perspective. The results of this paper evaluate the differences in how regulatory, safety basis, and operational practices must adapt to the dynamic environment of decontamination and decommissioning in contrast to the relatively constant

Investigation and assessment of lead slag concrete as nuclear shields

International Nuclear Information System (INIS)

Zaghloul, Y.R.

2009-01-01

The present work is concerned with the efficiency of heavy weight concrete as a shielding material in constructing nuclear installations as well as for radioactive wastes disposal facilities.In this context, lead slag was used as a replacement for fine aggregates in heavy concrete shields that include local heavy weight aggregates (namely; barite and ilmenite) as well as normal concrete includes dolomite and sand as coarse and fine aggregates, as a reference. The effect of different percentages of lead slag was investigated to assess the produced lead slag concrete as a nuclear shielding material. The different properties (physical, mechanical and nuclear) of the produced lead slag concrete were investigated. The results obtained showed that increasing the lead slag percentage improving the investigated properties of the different concrete mixes. In addition, ilmenite concrete with 20% lead slag showed the best results for all the investigated properties.

Re-utilization of concrete from decommissioned nuclear facilities by re-clinkering

International Nuclear Information System (INIS)

Costes, J.R.

1998-01-01

Usually concrete is reused as aggregates. To maximize reusing by producing a new concrete, separation of its fundamental components (aggregates, sand, cement) is required. But can already hydrated cement be re-clinkered and become new good cement again? This paper shows how specimen of pure cement paste of ordinary Portland cement was hydrated then crushed and reclinkerised. Chemical and, X ray diffraction analysis, electronic microscope coupled with X fluorescence observations, demonstrate that very few differences occurs in the cement paste before and after re-clinkering; mechanical properties of the reactivated powder are very satisfactory, even exceeding the strength of the original pure cement paste. As this process involves heating up to 1450 deg C, some of the contamination (caesium, ruthenium) may also be eliminated by volatilization and recovery on filters. This process could be very promising to complete recycle large quantities of low level contaminated concrete inside the nuclear industry as alternative to storage in specific and expensive dump, as it is now the case for steel materials. (authors)

PROSPECTS OF ESTABLISHING EARTHQUAKE RESISTANT BUILDINGS FROM TUBE CONCRETE CONSTRUCTIONS

Directory of Open Access Journals (Sweden)

Abdujafar I. Akaev

2017-01-01

Full Text Available Abstract. Objectives The aim of the research is to find optimal design solutions for the erection of buildings that will ensure their reliability and durability, compliance with environmental requirements, fire resistance and earthquake resistance. In this regard, the task is to determine the advantages and prospects of erecting earthquake resistant buildings from tube concrete constructions, since they are distinct in constructive, technological and economic efficiency when are used as vertical load-bearing struts of high-rise buildings. Method The technique for calculating the strength of normal sections of eccentrically-compressed tube concrete elements uses a nonlinear deformation model, taking into account the joint operation of the steel shell and the concrete core under the conditions of triaxial compression. Results In the article the review of the newest world experience of using tube concrete as vertical load-bearing structures for public facilities from the standpoint of earthquake resistant construction is given. The international practices of public facility construction ranging in height from 100 to 600 m with the use of tube concrete technology, including regions with dangerous natural and man-made conditions, have been studied. The structural, operational and technological advantages and disadvantages of tube concrete technology are analysed. Methods for calculating the strength of concrete tube elements in the case of central compression are considered: according to the so-called deformation theory, the state of total destruction of both concrete and tube fluidity attained at maximum pressure are indicated by the beginning of "tube flow on the longitudinal axis". The advantages and disadvantages of both methods are shown. Factors constraining the introduction and wider application of tube concrete constructions in Russia are considered. Conclusion While the advantages of concrete tube constructions in their extensive

Contingent post-closure plan, hazardous waste management units at selected maintenance facilities, US Army National Training Center, Fort Irwin, California

International Nuclear Information System (INIS)

1992-01-01

The National Training Center (NTC) at Fort Irwin, California, is a US Army training installation that provides tactical experience for battalion/task forces and squadrons in a mid- to high-intensity combat scenario. Through joint exercises with US Air Force and other services, the NTC also provides a data source for improvements of training doctrines, organization, and equipment. To meet the training and operational needs of the NTC, several maintenance facilities provide general and direct support for mechanical devices, equipment, and vehicles. Maintenance products used at these facilities include fuels, petroleum-based oils, lubricating grease, various degreasing solvents, antifreeze (ethylene glycol), transmission fluid, brake fluid, and hydraulic oil. Used or spent petroleum-based products generated at the maintenance facilities are temporarily accumulated in underground storage tanks (USTs), collected by the NTC hazardous waste management contractor (HAZCO), and stored at the Petroleum, Oil, and Lubricant (POL) Storage Facility, Building 630, until shipped off site to be recovered, reused, and/or reclaimed. Spent degreasing solvents and other hazardous wastes are containerized and stored on-base for up to 90 days at the NTC's Hazardous Waste Storage Facility, Building 703. The US Environmental Protection Agency (EPA) performed an inspection and reviewed the hazardous waste management operations of the NTC. Inspections indicated that the NTC had violated one or more requirements of Subtitle C of the Resource Conservation and Recovery Act (RCRA) and as a result of these violations was issued a Notice of Noncompliance, Notice of Necessity for Conference, and Proposed Compliance Schedule (NON) dated October 13, 1989. The following post-closure plan is the compliance-based approach for the NTC to respond to the regulatory violations cited in the NON

Design of buried concrete encasements

International Nuclear Information System (INIS)

Drake, R.M.

1989-01-01

The operation of many Department of Energy (DOE) sites requires the transfer of radioactive liquid products from one location to another. DOE Order 6430.1A requires that the transfer pipelines be designed and constructed so that any leakage can be detected and contained before it reaches the environment. One design option often considered to meet this requirement is to place the pipeline in a stainless steel-lined, buried concrete encasement. This provides the engineer with the design challenge to integrate standard structural design principles with unique DOE requirements. The complete design of a buried concrete encasement must consider seismic effects, leak detection, leak confinement, radiation shielding, thermal effects, pipe supports, and constructability. This paper contains a brief discussion of each of these design considerations, based on experience gained during the design of concrete encasements for the Process Facilities Modifications (PFM) project at Hanford

Properties of slag concrete for low-level waste containment

International Nuclear Information System (INIS)

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

1991-01-01

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

Evaluation of Geopolymer Concrete for Rocket Test Facility Flame Deflectors

Science.gov (United States)

Allgood, Daniel C.; Montes, Carlos; Islam, Rashedul; Allouche, Erez

2014-01-01

The current paper presents results from a combined research effort by Louisiana Tech University (LTU) and NASA Stennis Space Center (SSC) to develop a new alumina-silicate based cementitious binder capable of acting as a high performance refractory material with low heat ablation rate and high early mechanical strength. Such a binder would represent a significant contribution to NASA's efforts to develop a new generation of refractory 'hot face' liners for liquid or solid rocket plume environments. This project was developed as a continuation of on-going collaborations between LTU and SSC, where test sections of a formulation of high temperature geopolymer binder were cast in the floor and walls of Test Stand E-1 Cell 3, an active rocket engine test stand flame trench. Additionally, geopolymer concrete panels were tested using the NASA-SSC Diagnostic Test Facility (DTF) thruster, where supersonic plume environments were generated on a 1ft wide x 2ft long x 6 inch deep refractory panel. The DTF operates on LOX/GH2 propellants producing a nominal thrust of 1,200 lbf and the combustion chamber conditions are Pc=625psig, O/F=6.0. Data collected included high speed video of plume/panel area and surface profiles (depth) of the test panels measured on a 1-inch by 1-inch giving localized erosion rates during the test. Louisiana Tech conducted a microstructure analysis of the geopolymer binder after the testing program to identify phase changes in the material.

Treatment of Radioactive Contaminated Soil and Concrete Wastes Using the Regulatory Clearance

Energy Technology Data Exchange (ETDEWEB)

Kang, Il Sik; Ryu, W. S.; Kim, T. K.; Shon, J. S.; Ahn, S. J.; Lee, Y. H.; Bae, S. M.; Hong, D. S.; Ji, Y. Y.; Lee, B. C

2008-11-15

In the radioactive waste storage facilities at the Korea Atomic Energy Research Institute (KAERI) in Daejoen, there are thousands drums of radioactive contaminated soil and concrete wastes. The soil and concrete wastes were generated in 1988 during the decommissioning process of the research reactor and the attached radioactive waste treatment facility which were located in Seoul. The wastes were transported to Daejeon and have been stored since then. At the generation time, the radioactive contamination of the wastes was very low, and the radionuclides in the wastes was Co-60 and Cs-137. As the wastes have been stored for more than 20 years, the radioactivity concentration of the wastes has been decayed to become very extremely low. The wastes are needed to be treated because they take up large spaces at the storage facility. Also by treating the wastes, final disposal cost can be saved. So, the regulatory clearance was considered as a treatment method for the soil and concrete wastes with extremely low radioactivity concentration.

Use of sulfur concrete for radioecological problems solution in Kazakhstan

International Nuclear Information System (INIS)

Takibaev, Zh.; Belyashov, D.; Vagin, S.

2001-01-01

At present during intensive development of oil and gas fields in Kazakhstan a lot amount of sulfur is extracting. The problem of sulfur utilization demands its immediate solution. One of the perspective trends of sulfur utilization is use it in production of sulfur polymer concrete. It is well known, that encapsulation of low level radioactive and toxic wastes in sulfur polymer concrete and design from it radiation protection facilities have good perspectives for solution of radioecological problems. Sulfur concrete has high corrosion and radiation stability, improved mechanical and chemical properties. Unique properties of sulfur concrete allow to use it in materials ensuring protection from external irradiation

Innovative technologies for recycling contaminated concrete and scrap metal

International Nuclear Information System (INIS)

Bossart, S.J.; Moore, J.

1993-01-01

Decontamination and decommissioning of US DOE's surplus facilities will generate enormous quantities of concrete and scrap metal. A solicitation was issued, seeking innovative technologies for recycling and reusing these materials. Eight proposals were selected for award. If successfully developed, these technologies will enable DOE to clean its facilities by 2019

IDENTIFICATION OF DOE'S POST-CLOSURE MONITORING NEEDS AND REQUIREMENTS

Energy Technology Data Exchange (ETDEWEB)

M.A. Ebadian, Ph.D.

1999-01-01

The 2006 plan sets an ambitious agenda for the U.S. Department of Energy (DOE), Office of Environmental Management (EM) and the remediation of sites contaminated by decades of nuclear weapons production activities. The plan's primary objective is to reduce overall clean up costs by first eliminating the environmental problems that are most expensive to control and safely maintain. In the context of the 2006 Plan, closure refers to the completion of area or facility specific cleanup projects. The cleanup levels are determined by the planned future use of the site or facility. Use restrictions are still undecided for most sites but are highly probable to exclude residential or agricultural activities. Most of the land will be remediated to ''industrial use'' levels with access restrictions and some areas will be closed-off through containment. Portions of the site will be reserved for waste disposal, either as a waste repository or the in-situ immobilization of contaminated soil and groundwater, and land use will be restricted to waste disposal only. The land used for waste disposal will require monitoring and maintenance activities after closure. Most of the land used for industrial use may also require such postclosure activities. The required postclosure monitoring and maintenance activities will be imposed by regulators and stakeholders. Regulators will not approve closure plans without clearly defined monitoring methods using approved technologies. Therefore, among all other more costly and labor-intensive closure-related activities, inadequate planning for monitoring and lack of appropriate monitoring technologies can prevent closure. The purpose of this project is to determine, document, and track the current and evolving postclosure monitoring requirements at DOE-EM sites. This information will aid CMST-CP in guiding its postclosure technology development and deployment efforts.

HYDRAULIC AND PHYSICAL PROPERTIES OF SALTSTONE GROUTS AND VAULT CONCRETES

International Nuclear Information System (INIS)

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

2008-01-01

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

Risk and Performance Analyses Supporting Closure of WMA C at the Hanford Site in Southeast Washington

International Nuclear Information System (INIS)

Eberlein, Susan J.; Bergeron, Marcel P.; Kemp, Christopher J.; Hildebrand, R. Douglas; Aly, Alaa; Kozak, Matthew; Mehta, Sunil; Connelly, Michael

2013-01-01

The Office of River Protection under the U.S. Department of Energy (DOE) is pursuing closure of the Single-Shell Tank (SST) Waste Management Area (WMA) C as stipulated by the Hanford Federal Facility Agreement and Consent Order (HFFACO) under federal requirements and work tasks will be done under the State-approved closure plans and permits. An initial step in meeting the regulatory requirements is to develop a baseline risk assessment representing current conditions based on available characterization data and information collected at the WMA C location. The baseline risk assessment will be supporting a Resource Conservation and Recovery Act of 1976 (RCRA) Field Investigation (RFI)/Corrective Measures Study (CMS) for WMA closure and RCRA corrective action. Complying with the HFFACO conditions also involves developing a long-term closure Performance Assessment (PA) that evaluates human health and environmental impacts resulting from radionuclide inventories in residual wastes remaining in WMA C tanks and ancillary equipment. This PA is being developed to meet the requirements necessary for closure authorization under DOE Order 435.1 and Washington State Hazardous Waste Management Act. To meet the HFFACO conditions, the long-term closure risk analysis will include an evaluation of human health and environmental impacts from hazardous chemical inventories along with other performance Comprehensive Environmental Response, Compensation, and Liability Act Appropriate and Applicable Requirements (CERCLA ARARs) in residual wastes left in WMA C facilities after retrieval and removal. This closure risk analysis is needed to needed to comply with the requirements for permitted closure. Progress to date in developing a baseline risk assessment of WMA C has involved aspects of an evaluation of soil characterization and groundwater monitoring data collected as a part of the RFI/CMS and RCRA monitoring. Developing the long-term performance assessment aspects has involved the

Risk and Performance Analyses Supporting Closure of WMA C at the Hanford Site in Southeast Washington

Energy Technology Data Exchange (ETDEWEB)

Eberlein, Susan J.; Bergeron, Marcel P.; Kemp, Christopher J.

2013-11-11

The Office of River Protection under the U.S. Department of Energy (DOE) is pursuing closure of the Single-Shell Tank (SST) Waste Management Area (WMA) C as stipulated by the Hanford Federal Facility Agreement and Consent Order (HFFACO) under federal requirements and work tasks will be done under the State-approved closure plans and permits. An initial step in meeting the regulatory requirements is to develop a baseline risk assessment representing current conditions based on available characterization data and information collected at the WMA C location. The baseline risk assessment will be supporting a Resource Conservation and Recovery Act of 1976 (RCRA) Field Investigation (RFI)/Corrective Measures Study (CMS) for WMA closure and RCRA corrective action. Complying with the HFFACO conditions also involves developing a long-term closure Performance Assessment (PA) that evaluates human health and environmental impacts resulting from radionuclide inventories in residual wastes remaining in WMA C tanks and ancillary equipment. This PA is being developed to meet the requirements necessary for closure authorization under DOE Order 435.1 and Washington State Hazardous Waste Management Act. To meet the HFFACO conditions, the long-term closure risk analysis will include an evaluation of human health and environmental impacts from hazardous chemical inventories along with other performance Comprehensive Environmental Response, Compensation, and Liability Act Appropriate and Applicable Requirements (CERCLA ARARs) in residual wastes left in WMA C facilities after retrieval and removal. This closure risk analysis is needed to needed to comply with the requirements for permitted closure. Progress to date in developing a baseline risk assessment of WMA C has involved aspects of an evaluation of soil characterization and groundwater monitoring data collected as a part of the RFI/CMS and RCRA monitoring. Developing the long-term performance assessment aspects has involved the

221-U Facility concrete and reinforcing steel evaluations specification for the canyon disposition initiative (CDI)

International Nuclear Information System (INIS)

Baxter, J.T.

1998-01-01

This describes a test program to establish the in-situ material properties of the reinforced concrete in Building 221-U for comparison to the original design specifications. Field sampling and laboratory testing of concrete and reinforcing steel structural materials in Building 221-U for design verification will be undertaken. Forty seven samples are to be taken from radiologically clean exterior walls of the canyon. Laboratory testing program includes unconfined compressive strength of concrete cores, tensile strength of reinforcing steel, and petrographic examinations of concrete cores taken from walls below existing grade

Village facilities and social place attachment in the rural Netherlands

NARCIS (Netherlands)

Gieling, J.; Haartsen, T.; Vermeij, Lotte; Svels, Kristina

Economies of scale and increased mobility have led to the closure of many village facilities. Most residents do not rely on locally available facilities anymore for their primary function. However, facilities are also meeting places. A decline in facilities may therefore negatively influence

Concrete decontamination by Electro-Hydraulic Scabbling (EHS). Topical report

International Nuclear Information System (INIS)

1996-01-01

Electro-Hydraulic Scabbling (EHS) technology and equipment for decontaminating concrete structures from radionuclides, organic substances, and hazardous metals is being developed by Textron Systems Division (TSD). This wet scabbling technique involves the generation of powerful shock waves and intense cavitation by a strong pulsed electric discharge in a water layer at the concrete surface. The high pressure impulse results in stresses which crack and peel off a concrete layer of a controllable thickness. Scabbling produces contaminated debris of relatively small volume which can be easily removed, leaving clean bulk concrete. This new technology is being developed under Contract No. DE-AC21-93MC30164. The project objective is to develop and demonstrate a cost-efficient, rapid, controllable process to remove the surface layer of contaminated concrete while generating minimal secondary waste. The primary target of this program is uranium-contaminated concrete floors which constitute a substantial part of the contaminated area at DOE weapon facilities

218-E-8 Borrow Pit Demolition Site closure plan. Revision 1

International Nuclear Information System (INIS)

Ruck, F.R.

1994-01-01

The 218-E-8 Demolition Site was the site of a single demolition event in November of 1984. This demolition event was a form of thermal treatment for discarded explosive chemical products. Because the 218-E-8 Demolition Site will no longer be used for this thermal activity, the site will be closed. Closure will be conducted pursuant to the requirements of the Washington State Department of Ecology (Ecology) ''Dangerous Waste Regulations,'' Washington Administrative Code (WAC) 173-303-610 and 40 Code of Federal Regulations (CFR) 270.1. The 218-E-8 Borrow Pit Demolition Site Closure Plan consists of a Hanford Facility Dangerous Waste Part A Permit Application, Form 3, Revision 4, and a closure plan. An explanation of the Part A Form 3, submitted with this closure plan is provided at the beginning of the Part A Section. The closure plan consists of nine chapters and five appendices. This closure plan presents a description of the 218-E-8 Demolition Site, the history of the waste treated, and the approach that will be followed to close the 218-E-8 Demolition Site. Because there were no radioactively contaminated chemicals involved in t he demolitions at the 218-E-8 Borrow Pit site, the information on radionuclides is provided for ''information only.'' Remediation of any radioactive contamination is not within the scope of this closure plan. Only dangerous constituents derived from 218-E-8 Demolition Site operations will be addressed in this closure plan in accordance with WAC 173-303-610(2)(b)(i)

Explosive Spalling of Fire Exposed Resource Saving Concrete Structures

DEFF Research Database (Denmark)

Sørensen, Lars Schiøtt; Hertz, Kristian Dahl; Kristiansen, Finn Harken

2003-01-01

The paper describes briefly a new test facility, which has been developed within the project “Resource Saving Concrete Structures”, also called “Green Concrete” and some test results from the project. A full report is available from the home page of the Department of Civil Engineering Kristiansen......, Hertz, Sørensen [1]. The main idea was to establish a test method by means of which it should be possible to assess whether a particular concrete has an increased risk of spalling compared to traditional concretes as defined in Hertz [2] and only using ordinary standard cylinders as test specimens....... The method has been applied on the green concretes of the project and later also as a first indicator in other projects. The method appears to be a valuable tool for the first investigation of new concretes...

Two-dimensional interaction of oxidic corium with concretes: The VULCANO VB test series

Energy Technology Data Exchange (ETDEWEB)

Journeau, Christophe [CEA, DEN, STRI/LMA, Cadarache, F-13108 St Paul lez Durance (France)], E-mail: christophe.journeau@cea.fr; Piluso, Pascal; Haquet, Jean-Francois; Boccaccio, Eric; Saldo, Valerie; Bonnet, Jean-Michel; Malaval, Sophie; Carenini, Laure [CEA, DEN, STRI/LMA, Cadarache, F-13108 St Paul lez Durance (France); Brissonneau, Laurent [CEA, DEN, STPA/LPC, Cadarache, F-13108 St Paul lez Durance (France)

2009-10-15

Three two-dimensional Molten Core-Concrete Interaction tests have been conducted in the VULCANO facility with prototypic oxidic corium. The major finding is that for the two tests with silica-rich concrete, the ablation was anisotropic while it was isotropic for limestone-rich concrete. The cause of this behaviour is not yet well understood. Post Test Examinations have indicated that for the silica-rich concrete, the corium melt mixed specifically with mortar, while, for limestone-rich concretes, the analysed samples were in accordance with a corium-concrete mixing. The experimental results are described and compared to numerical codes. Separate Effect Tests with Artificial Concretes and prototypic corium are proposed to understand the phenomena governing the ablation geometry.

Two-dimensional interaction of oxidic corium with concretes: The VULCANO VB test series

International Nuclear Information System (INIS)

Journeau, Christophe; Piluso, Pascal; Haquet, Jean-Francois; Boccaccio, Eric; Saldo, Valerie; Bonnet, Jean-Michel; Malaval, Sophie; Carenini, Laure; Brissonneau, Laurent

2009-01-01

Three two-dimensional Molten Core-Concrete Interaction tests have been conducted in the VULCANO facility with prototypic oxidic corium. The major finding is that for the two tests with silica-rich concrete, the ablation was anisotropic while it was isotropic for limestone-rich concrete. The cause of this behaviour is not yet well understood. Post Test Examinations have indicated that for the silica-rich concrete, the corium melt mixed specifically with mortar, while, for limestone-rich concretes, the analysed samples were in accordance with a corium-concrete mixing. The experimental results are described and compared to numerical codes. Separate Effect Tests with Artificial Concretes and prototypic corium are proposed to understand the phenomena governing the ablation geometry.

In situ treatment of concrete surfaces by organic impregnation and polymerization

International Nuclear Information System (INIS)

Ursella, P.; Moretti, G.; Pellecchia, V.

1990-01-01

The impregnation by resins of concrete structures is a process well known at PIC (Polymer Impregnated Concrete). This process improves the physical-chemical features of concrete matrixes in order to extend their durability when severe environmental conditions may occur. The main objective of this research contract has been the verification of a proper impregnation 'in situ' of existing concrete surfaces, of any laying in the space, by means of a prototype machine, expressly designed and implemented, and verification of the increase of mechanical resistance, leach resistance, durability of treated material. In a nuclear facility this goal is very important in relation to the long term integrity of concrete structures during operating lifetime and, in particular, after final shutdown. (author)

Quality Control of Concrete Structure For APR1400 Construction

International Nuclear Information System (INIS)

Seo, Inseop; Song, Changhak; Kim, Duill

2012-01-01

Nuclear structure shall be constructed to protect internal facilities in the normal operation against external accidents such as the radiation shielding, earthquakes and to be leak-proof of radioactive substances to the external environment in case of loss of coolants. containment and auxiliary building of nuclear power plants are built in reinforced concrete structures to maintain these protection functions. Nuclear structures shall be designed to ensure soundness in operation since they are located on the waterfront where is easy do drain the cooling water and so deterioration and damage of concrete structures caused by seawater can occur. Durability is ensured for concrete structures of APR1400, a Korea standard NPP, in compliance with all safety requirements. In particular, owners perform quality control directly on the production and pouring of cast in place concrete for the concrete structure construction to make sure concrete structures established with quality homogeneity and durability. This report is to look into the quality control standard and management status of cast in place concrete for APR1400 construction

An Analysis of Radiation Penetration through the U-Shaped Cast Concrete Joints of Concrete Shielding in the Multipurpose Gamma Irradiator of BATAN

Science.gov (United States)

Ardiyati, Tanti; Rozali, Bang; Kasmudin

2018-02-01

An analysis of radiation penetration through the U-shaped joints of cast concrete shielding in BATAN’s multipurpose gamma irradiator has been carried out. The analysis has been performed by calculating the radiation penetration through the U-shaped joints of the concrete shielding using MCNP computer code. The U-shaped joints were a new design in massive concrete construction in Indonesia and, in its actual application, it is joined by a bonding agent. In the MCNP simulation model, eight detectors were located close to the observed irradiation room walls of the concrete shielding. The simulation results indicated that the radiation levels outside the concrete shielding was less than the permissible limit of 2.5 μSv/h so that the workers could safely access electrical room, control room, water treatment facility and outside irradiation room. The radiation penetration decreased as the density of material increased.

Material balance and diet in bioregenerative life support systems: Connection with coefficient of closure

Science.gov (United States)

Manukovsky, N. S.; Kovalev, V. S.; Somova, L. A.; Gurevich, Yu. L.; Sadovsky, M. G.

Bioregenerative life support systems (BLSS) with different coefficients of closure are considered. The 66.2% coefficient of closure achieved in "BIOS-3" facility experiments has been taken as a base value. The increase in coefficient of closure up to 72.6-93.0% is planned due to use of soil-like substrate (SLS) and concentrating of urine. Food values were estimated both in a base variant ("BIOS-3"), and with increases in the coefficient of closure. It is shown that food requirements will be more fully satisfied by internal crop production with an increase in the coefficient of closure of the BLSS. Changes of massflow rates on an 'input-output' and inside BLSS are considered. Equations of synthesis and degradation of organic substances in BLSS were examined using a stoichiometric model. The paper shows that at incomplete closure of BLSS containing SLS there is a problem of nitrogen balancing. To compensate for the removal of nitrogen from the system in urine and feces, it is necessary to introduce food and a nitrogen-containing additive.

Transportable Vitrification System RCRA Closure Practical Waste Disposition Saves Time And Money

International Nuclear Information System (INIS)

Brill, Angie; Boles, Roger; Byars, Woody

2003-01-01

The Transportable Vitrification System (TVS) was a large-scale vitrification system for the treatment of mixed wastes. The wastes contained both hazardous and radioactive materials in the form of sludge, soil, and ash. The TVS was developed to be moved to various United States Department of Energy (DOE) facilities to vitrify mixed waste as needed. The TVS consists of four primary modules: (1) Waste and Additive Materials Processing Module; (2) Melter Module; (3) Emissions Control Module; and (4) Control and Services Module. The TVS was demonstrated at the East Tennessee Technology Park (ETTP) during September and October of 1997. During this period, approximately 16,000 pounds of actual mixed waste was processed, producing over 17,000 pounds of glass. After the demonstration was complete it was determined that it was more expensive to use the TVS unit to treat and dispose of mixed waste than to direct bury this waste in Utah permitted facility. Thus, DOE had to perform a Resource Conservation and Recovery Act (RCRA) closure of the facility and find a reuse for as much of the equipment as possible. This paper will focus on the following items associated with this successful RCRA closure project: TVS site closure design and implementation; characterization activities focused on waste disposition; pollution prevention through reuse; waste minimization efforts to reduce mixed waste to be disposed; and lessons learned that would be integrated in future projects of this magnitude

Radionuclide Retention in Concrete Wasteforms

Energy Technology Data Exchange (ETDEWEB)

Wellman, Dawn M.; Jansik, Danielle P.; Golovich, Elizabeth C.; Cordova, Elsa A.

2012-09-24

Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation); the mechanism of contaminant release; the significance of contaminant release pathways; how wasteform performance is affected by the full range of environmental conditions within the disposal facility; the process of wasteform aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of wasteform aging on chemical, physical, and radiological properties; and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the wasteforms come in contact with groundwater. Data collected throughout the course of this work will be used to quantify the efficacy of concrete wasteforms, similar to those used in the disposal of LLW and MLLW, for the immobilization of key radionuclides (i.e., uranium, technetium, and iodine). Data collected will also be used to quantify the physical and chemical properties of the concrete affecting radionuclide retention.

Overview of ORNL/NRC programs addressing durability of concrete structures

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.

1994-01-01

The role of reinforced concrete relative to its applications as either safety-related structures in nuclear power or engineered barriers of low-level radioactive waste disposal facilities is described. Factors that can affect the long-term durability of reinforced concrete are identified. Overviews are presented of the Structural Aging Program, which is addressing the aging management of safety-related concrete structures in nuclear power plants, and the Permeability Test Methods and Data Program, which is identifying pertinent data and information for use in performance assessments of engineered barriers for low-level radioactive waste disposal

Neutron shielding properties of a new high-density concrete

International Nuclear Information System (INIS)

Lorente, A.; Gallego, E.; Vega Carrillo, H.R.; Mendez, R.

2008-01-01

The neutron shielding properties of a new high-density concrete (commercially available under the name Hormirad TM , developed in Spain by the company CT-RAD) have been characterized both experimentally and by Monte Carlo calculations. The shielding properties of this concrete against photons were previously studied and the material is being used to build bunkers, mazes and doors in medical accelerator facilities with good overall results. In this work, the objective was to characterize the material behaviour against neutrons, as well as to test alternative mixings including boron compounds in an effort to improve neutron shielding efficiency. With that purpose, Hormirad TM slabs of different thicknesses were exposed to an 241 Am-Be neutron source under controlled conditions in the neutron measurements laboratory of the Nuclear Engineering Department at UPM. The original mix, which includes a high fraction of magnetite, was then modified by adding different proportions of anhydrous borax (Na 2 B 4 O 7 ). In order to have a reference against common concrete used to shield medical accelerator facilities, the same experiment was repeated with ordinary (HA-25) concrete slabs. In parallel to the experiments, Monte Carlo calculations of the experiments were performed with MCNP5. The experimental results agree reasonably well with the Monte Carlo calculations. Therefore, the first and equilibrium tenth-value layers have been determined for the different types of concrete tested. The results show an advantageous behaviour of the Hormirad TM concrete, in terms of neutron attenuation against real thickness of the shielding. Borated concretes seem less practical since they did not show better neutron attenuation with respect to real thickness and their structural properties are worse. The neutron attenuation properties of Hormirad TM for typical neutron spectra in clinical LINAC accelerators rooms have been also characterized by Monte Carlo calculation. (author)

Development of Practical Remediation Process for Uranium-Contaminated Concrete

Energy Technology Data Exchange (ETDEWEB)

Kim, S. S.; Kim, W. S.; Kim, G. N.; Moon, J. K. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

A volume reduction of the concrete waste by the appropriate treatment technologies will decrease the amount of waste to be disposed of and result in a reduction of the disposal cost and an enhancement of the efficiency of the disposal site. Our group has developed a 100 drums/year decontamination process and facilities for the decontamination of radioactive concrete. This practical scale process is little known. A practical decontamination process was developed to remove uranium from concrete pieces generated from the decommissioning of a uranium conversion plant. The concrete pieces are divided into two groups: concrete coated with and without epoxy. For the removal of epoxy from the concrete, direct burning by an oil flame is preferable to an electric heating method. The concrete blocks are crushed to below 30 mm and sifted to 1 mm. When the concrete pieces larger than 1 mm are sequentially washed with a clear washing solution and 1.0 M of nitric acid, most of their radioactivity reaches below the limit value of uranium for self-disposal. The concrete pieces smaller than 1 mm are decontaminated in a rotary washing machine by nitric acid, and an electrokinetic equipment is also used if their radioactivity is high.

Development of Practical Remediation Process for Uranium-Contaminated Concrete

International Nuclear Information System (INIS)

Kim, S. S.; Kim, W. S.; Kim, G. N.; Moon, J. K.

2013-01-01

A volume reduction of the concrete waste by the appropriate treatment technologies will decrease the amount of waste to be disposed of and result in a reduction of the disposal cost and an enhancement of the efficiency of the disposal site. Our group has developed a 100 drums/year decontamination process and facilities for the decontamination of radioactive concrete. This practical scale process is little known. A practical decontamination process was developed to remove uranium from concrete pieces generated from the decommissioning of a uranium conversion plant. The concrete pieces are divided into two groups: concrete coated with and without epoxy. For the removal of epoxy from the concrete, direct burning by an oil flame is preferable to an electric heating method. The concrete blocks are crushed to below 30 mm and sifted to 1 mm. When the concrete pieces larger than 1 mm are sequentially washed with a clear washing solution and 1.0 M of nitric acid, most of their radioactivity reaches below the limit value of uranium for self-disposal. The concrete pieces smaller than 1 mm are decontaminated in a rotary washing machine by nitric acid, and an electrokinetic equipment is also used if their radioactivity is high

Decommissioning of the Risoe hot cell facility

International Nuclear Information System (INIS)

Carlsen, H.

1992-02-01

Concise descriptions of actions taken in relation to the decommissioning of the hot cell facility at Risoe National Laboratory are presented. The removal of fissile material, of large contaminated equipment from the concrete cell line and a separate shielded storage facility, and the removal of large contaminated facilities such as out cell parts of a tube transport system between a concrete cell and a lead shielded steel box and a remotely operated Reichert Telatom microscope housed in a lead shielded glove box is described in addition to the initial mapping of radiation levels related to the decontamination of concrete cells. The dose commitment of 17.7 mSv was ascribed to 12 persons in the 2nd half of 1991. The work resulting in these doses was mainly handling of waste together with the frogman entrances in order to repair the in-cell crane and power manipulator. The overall time schedule for the project still appears to be applicable. (AB)

Overview of experimental results obtained under the Prestressed Concrete Nuclear Pressure Vessel Development Program at the Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Naus, D.J.

1978-01-01

Under the Prestressed Concrete Nuclear Pressure Vessel Development Program at the Oak Ridge National Laboratory, various aspects of Prestressed Concrete Pressure Vessels (PCPVs) are investigated and evaluated with respect to reliability, structural performance, constructability, and economy. Based upon identified needs, analytical and experimental investigations are conducted. Areas of interest include finite-element analysis development, materials and structural behavior tests, instrumentation evaluation and development, and structural model tests. Studies have been recently completed in the following areas: concrete embedment instrumentation systems for PCPVs, grouted-nongrouted prestressing systems, acoustic emission as a technique for structural integrity monitoring, and model tests of steam-generator cavity closure plugs for a Gas-Cooled Fast Reactor (GCFR). An overview of results is presented

Concrete Waste Recycling Process for High Quality Aggregate

International Nuclear Information System (INIS)

Ishikura, Takeshi; Fujii, Shin-ichi

2008-01-01

Large amount of concrete waste generates during nuclear power plant (NPP) dismantling. Non-contaminated concrete waste is assumed to be disposed in a landfill site, but that will not be the solution especially in the future, because of decreasing tendency of the site availability and natural resources. Concerning concrete recycling, demand for roadbeds and backfill tends to be less than the amount of dismantled concrete generated in a single rural site, and conventional recycled aggregate is limited of its use to non-structural concrete, because of its inferior quality to ordinary natural aggregate. Therefore, it is vital to develop high quality recycled aggregate for general uses of dismantled concrete. If recycled aggregate is available for high structural concrete, the dismantling concrete is recyclable as aggregate for industry including nuclear field. Authors developed techniques on high quality aggregate reclamation for large amount of concrete generated during NPP decommissioning. Concrete of NPP buildings has good features for recycling aggregate; large quantity of high quality aggregate from same origin, record keeping of the aggregate origin, and little impurities in dismantled concrete such as wood and plastics. The target of recycled aggregate in this development is to meet the quality criteria for NPP concrete as prescribed in JASS 5N 'Specification for Nuclear Power Facility Reinforced Concrete' and JASS 5 'Specification for Reinforced Concrete Work'. The target of recycled aggregate concrete is to be comparable performance with ordinary aggregate concrete. The high quality recycled aggregate production techniques are assumed to apply for recycling for large amount of non-contaminated concrete. These techniques can also be applied for slightly contaminated concrete dismantled from radiological control area (RCA), together with free release survey. In conclusion: a technology on dismantled concrete recycling for high quality aggregate was developed

Manufacture and quality control of concrete for Ikata Nuclear Power Station, Shikoku Electric Power Co., Inc

International Nuclear Information System (INIS)

Tada, Akiomi; Kitada, Takao

1989-01-01

Ikata Nuclear Power Station, only one nuclear power station in Shikoku, is located at the root of Sada Peninsula on Seto Inland sea side. At present, No.1 and No.2 plants of 566 MW each are in commercial operation, and on the east side, No.3 plant is under construction. No.3 plant is a PWR type plant of 890 MWe output, and the start of commercial operation is scheduled in March, 1995. In the construction of No.3 plant, the concrete used for civil engineering and building works is about 430,000 m 3 , and for the improvement of the quality control of concrete, the pursuit of economic efficiency, the fostering of concrete technology of employee and the coprosperity with local industries, the facilities for manufacturing concrete were constructed within the premise of the power station. The amount of use of concrete and respective materials classified by respective fiscal years, and the amount of manufacture of concrete that determines the scale of the concrete plant are shown. As to the construction of the concrete plant, the foundation work was started in March, 1987, and the machine foundation and building works were started in May, 1987. The acceptance was completed on August 17, 1987. The facilities of manufacturing concrete, the manufacture of concrete, and the quality control of materials and concrete are reported. (author)

Durability aspects of high-performance concretes for a waste repository. Appendix 3: Canada

International Nuclear Information System (INIS)

Philipose, K.E.

2001-01-01

The IRUS facility for the disposal of low level radioactive waste at the Chalk River Nuclear Laboratories in Ontario, Canada relies on the durability of concrete for the required 500 years of service life. A research programme based on laboratory testing to design a durable concrete and assess its long-term behaviour was initiated in 1988. This appendix discusses the methodology to assess the long-term behaviour of concrete, and some initial observations. Longevity predictions for concrete formulations based on diffusion testing are also presented

324 Building radiochemical engineering cells, high-level vault, low-level vault, and associated areas closure plan

International Nuclear Information System (INIS)

Barnett, J.M.

1998-01-01

The Hanford Site, located adjacent to and north of Richland, Washington, is operated by the US Department of Energy, Richland Operations Office (RL). The 324 Building is located in the 300 Area of the Hanford Site. The 324 Building was constructed in the 1960s to support materials and chemical process research and development activities ranging from laboratory/bench-scale studies to full engineering-scale pilot plant demonstrations. In the mid-1990s, it was determined that dangerous waste and waste residues were being stored for greater than 90 days in the 324 Building Radiochemical Engineering Cells (REC) and in the High-Level Vault/Low-Level Vault (HLV/LLV) tanks. [These areas are not Resource Conservation and Recovery Act of 1976 (RCRA) permitted portions of the 324 Building.] Through the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-89, agreement was reached to close the nonpermitted RCRA unit in the 324 Building. This closure plan, managed under TPA Milestone M-20-55, addresses the identified building areas targeted by the Tri-Party Agreement and provides commitments to achieve the highest degree of compliance practicable, given the special technical difficulties of managing mixed waste that contains high-activity radioactive materials, and the physical limitations of working remotely in the areas within the subject closure unit. This closure plan is divided into nine chapters. Chapter 1.0 provides the introduction, historical perspective, 324 Building history and current mission, and the regulatory basis and strategy for managing the closure unit. Chapters 2.0, 3.0, 4.0, and 5.0 discuss the detailed facility description, process information, waste characteristics, and groundwater monitoring respectively. Chapter 6.0 deals with the closure strategy and performance standard, including the closure activities for the B-Cell, D-Cell, HLV, LLV; piping and miscellaneous associated building areas. Chapter 7.0 addresses the

Lining facility for FBR type reactor

International Nuclear Information System (INIS)

Shimano, Kunio.

1991-01-01

In a lining facility for protecting structural material concretes for concrete buildings in an FBR type power plant, sodium-resistant and heat-resistant first and second coating layers are lined at the surface of concretes, and steam releasing materials are disposed between the first and the second coating layers for releasing water contents evaporated from the concretes to the outside. With such a constitution, since there is no structures for welding steel plates to each other as in the prior art, the fabrication is made easy. Further, since cracks of coating materials can be suppressed, reactor safety is improved. (T.M.)

Risk management considerations for seismic upgrading of an older facility for short-term residue stabilization

International Nuclear Information System (INIS)

Additon, S.L.; Peregoy, W.L.; Foppe, T.L.

1999-01-01

Building 707 and its addition, Building 707A, were selected, after the production mission of Rocky Flats was terminated a few years ago, to stabilize many of the plutonium residues remaining at the site by 2002. The facility had undergone substantial safety improvements to its safety systems and conduct of operations for resumption of plutonium operations in the early 1990s and appeared ideally suited for this new mission to support accelerated Site closure. During development of a new authorization basis, a seismic evaluation was performed. This evaluation addressed an unanalyzed expansion joint and suspect connection details for the precast concrete tilt-up construction and concluded that the seismic capacity of the facility is less than half of that determined by previous analysis. Further, potential seismic interaction was identified between a collapsing Building 707 and the seismically upgraded Building 707A, possibly causing the partial collapse of the latter. Both the operating contractor and the Department of Energy sought a sound technical basis for deciding how to proceed. This paper addresses the risks of the as-is facility and possible benefits of upgrades to support a decision on whether to upgrade the seismic capacity of Building 707, accept the risk of the as-is facility for its short remaining mission, or relocate critical stabilization missions. The paper also addresses the Department of Energy's policy on natural phenomena

Geophysical Research Facility

Data.gov (United States)

Federal Laboratory Consortium — The Geophysical Research Facility (GRF) is a 60 ft long × 22 ft wide × 7 ft deep concrete basin at CRREL for fresh or saltwater investigations and can be temperature...

SRS tank closure. Innovative technology summary report

International Nuclear Information System (INIS)

1999-08-01

High-level waste (HLW) tank closure technology is designed to stabilize any remaining radionuclides and hazardous constituents left in a tank after bulk waste removal. Two Savannah River Site (SRS) HLW tanks were closed after cleansing and then filling each tank with three layers of grout. The first layer consists of a chemically reducing grout. The fill material has chemical properties that retard the movement of some radionuclides and chemical constituents. A layer of controlled low-strength material (CLSM), a self-leveling fill material, is placed on top of the reducing grout. CLSM provides sufficient strength to support the overbearing weight. The final layer is a free-flowing, strong grout similar to normal concrete. After the main tank cavity is filled, risers are filled with grout, and all waste transfer piping connected to the tank is isolated. The tank ventilation system is dismantled, and the remaining systems are isolated. Equipment that remains with the tank is filled with grout. The tank and ancillary systems are left in a state requiring only limited surveillance. Administrative procedures are in place to control land use and access. DOE eventually plans to remove all of its HLW storage tanks from service. These tanks are located at SRS, Hanford, and Idaho National Engineering and Environmental Laboratory. Low-activity waste storage tanks at Oak Ridge Reservation are also scheduled for closure

Properties of concrete blocks prepared with low grade recycled aggregates.

Science.gov (United States)

Poon, Chi-Sun; Kou, Shi-cong; Wan, Hui-wen; Etxeberria, Miren

2009-08-01

Low grade recycled aggregates obtained from a construction waste sorting facility were tested to assess the feasibility of using these in the production of concrete blocks. The characteristics of the sorted construction waste are significantly different from that of crushed concrete rubbles that are mostly derived from demolition waste streams. This is due to the presence of higher percentages of non-concrete components (e.g. >10% soil, brick, tiles etc.) in the sorted construction waste. In the study reported in this paper, three series of concrete block mixtures were prepared by using the low grade recycled aggregates to replace (i) natural coarse granite (10mm), and (ii) 0, 25, 50, 75 and 100% replacement levels of crushed stone fine (crushed natural granite concrete blocks. Test results on properties such as density, compressive strength, transverse strength and drying shrinkage as well as strength reduction after exposure to 800 degrees C are presented below. The results show that the soil content in the recycled fine aggregate was an important factor in affecting the properties of the blocks produced and the mechanical strength deceased with increasing low grade recycled fine aggregate content. But the higher soil content in the recycled aggregates reduced the reduction of compressive strength of the blocks after exposure to high temperature due probably to the formation of a new crystalline phase. The results show that the low grade recycled aggregates obtained from the construction waste sorting facility has potential to be used as aggregates for making non-structural pre-cast concrete blocks.

Dynamic behavior of reinforced concrete beam subjected to impact load

International Nuclear Information System (INIS)

Ito, Chihiro; Ohnuma, Hiroshi; Sato, Koichi; Takano, Hiroshi

1984-01-01

The purpose of this report is to find out the impact behavior of reinforced concrete beams by means of experiment. The reinforced concrete is widely used for such an important structure as the building facilities of the nuclear power plant, and so the impact behavior of the reinforced concrete structures must be examined to estimate the resistance of concrete containment against impact load and to develope the reasonable and reliable design procedure. The impact test on reinforced concrete beam which is one of the most basic elements in the structure was conducted. Main results are summarized as follows. 1) Bending failure occured on static test. On the other hand, shear failure occured in the case of high impact velocity on impact test. 2) Penetration depth and residual deflection are approximately proportional to V 2 (V: velocity at impact). 3) Flexural wave propagates about at the speed of 2000 m/s. 4) The resistance of reinforced concrete beam against the impact load is fairly good. (author)

DECONTAMINATION TECHNOLOGIES FOR FACILITY REUSE

International Nuclear Information System (INIS)

Bossart, Steven J.; Blair, Danielle M.

2003-01-01

As nuclear research and production facilities across the U.S. Department of Energy (DOE) nuclear weapons complex are slated for deactivation and decommissioning (D and D), there is a need to decontaminate some facilities for reuse for another mission or continued use for the same mission. Improved technologies available in the commercial sector and tested by the DOE can help solve the DOE's decontamination problems. Decontamination technologies include mechanical methods, such as shaving, scabbling, and blasting; application of chemicals; biological methods; and electrochemical techniques. Materials to be decontaminated are primarily concrete or metal. Concrete materials include walls, floors, ceilings, bio-shields, and fuel pools. Metallic materials include structural steel, valves, pipes, gloveboxes, reactors, and other equipment. Porous materials such as concrete can be contaminated throughout their structure, although contamination in concrete normally resides in the top quarter-inch below the surface. Metals are normally only contaminated on the surface. Contamination includes a variety of alpha, beta, and gamma-emitting radionuclides and can sometimes include heavy metals and organic contamination regulated by the Resource Conservation and Recovery Act (RCRA). This paper describes several advanced mechanical, chemical, and other methods to decontaminate structures, equipment, and materials

Demonstration recommendations for accelerated testing of concrete decontamination methods

Energy Technology Data Exchange (ETDEWEB)

Dickerson, K.S.; Ally, M.R.; Brown, C.H.; Morris, M.I.; Wilson-Nichols, M.J.

1995-12-01

A large number of aging US Department of Energy (DOE) surplus facilities located throughout the US require deactivation, decontamination, and decommissioning. Although several technologies are available commercially for concrete decontamination, emerging technologies with potential to reduce secondary waste and minimize the impact and risk to workers and the environment are needed. In response to these needs, the Accelerated Testing of Concrete Decontamination Methods project team described the nature and extent of contaminated concrete within the DOE complex and identified applicable emerging technologies. Existing information used to describe the nature and extent of contaminated concrete indicates that the most frequently occurring radiological contaminants are {sup 137}Cs, {sup 238}U (and its daughters), {sup 60}Co, {sup 90}Sr, and tritium. The total area of radionuclide-contaminated concrete within the DOE complex is estimated to be in the range of 7.9 {times} 10{sup 8} ft{sup 2}or approximately 18,000 acres. Concrete decontamination problems were matched with emerging technologies to recommend demonstrations considered to provide the most benefit to decontamination of concrete within the DOE complex. Emerging technologies with the most potential benefit were biological decontamination, electro-hydraulic scabbling, electrokinetics, and microwave scabbling.

Demonstration recommendations for accelerated testing of concrete decontamination methods

International Nuclear Information System (INIS)

Dickerson, K.S.; Ally, M.R.; Brown, C.H.; Morris, M.I.; Wilson-Nichols, M.J.

1995-12-01

A large number of aging US Department of Energy (DOE) surplus facilities located throughout the US require deactivation, decontamination, and decommissioning. Although several technologies are available commercially for concrete decontamination, emerging technologies with potential to reduce secondary waste and minimize the impact and risk to workers and the environment are needed. In response to these needs, the Accelerated Testing of Concrete Decontamination Methods project team described the nature and extent of contaminated concrete within the DOE complex and identified applicable emerging technologies. Existing information used to describe the nature and extent of contaminated concrete indicates that the most frequently occurring radiological contaminants are 137 Cs, 238 U (and its daughters), 60 Co, 90 Sr, and tritium. The total area of radionuclide-contaminated concrete within the DOE complex is estimated to be in the range of 7.9 x 10 8 ft 2 or approximately 18,000 acres. Concrete decontamination problems were matched with emerging technologies to recommend demonstrations considered to provide the most benefit to decontamination of concrete within the DOE complex. Emerging technologies with the most potential benefit were biological decontamination, electro-hydraulic scabbling, electrokinetics, and microwave scabbling

Bulk Building Material Characterization and Decontamination Using a Concrete Floor and Wall Contamination Profiling Technology

International Nuclear Information System (INIS)

Aggarwal, S.; Charters, G.; Blauvelt, D.

2002-01-01

The concrete profiling technology, RadPro(trademark) has four major components: a drill with a specialized cutting and sampling head, drill bits, a sample collection unit and a vacuum pump. The equipment in conjunction with portable radiometric instrumentation produces a profile of radiological or chemical contamination through the material being studied. The drill head is used under hammer action to penetrate hard surfaces. This causes the bulk material to be pulverized as the drill travels through the radioactive media efficiently transmitting to the sampling unit a representative sample of powdered bulk material. The profiling equipment is designed to sequentially collect all material from the hole. The bulk material samples are continuously retrieved by use of a specially designed vacuumed sample retrieval unit that prevents cross contamination of the clean retrieved samples. No circulation medium is required with this profiling process; therefore, the only by-product from drilling is the sample. The data quality, quantity, and representativeness may be used to produce an activity profile from the hot spot surface into the bulk building material. The activity data obtained during the profiling process is reduced and transferred to building drawings as part of a detailed report of the radiological problem. This activity profile may then be expanded to ultimately characterize the facility and expedite waste segregation and facility closure at a reduced cost and risk

Analysis of the status of pre-release cracks in prestressed concrete structures using long-gauge sensors

International Nuclear Information System (INIS)

Abdel-Jaber, H; Glisic, B

2015-01-01

Prestressed structures experience limited tensile stresses in concrete, which limits or completely eliminates the occurrence of cracks. However, in some cases, large tensile stresses can develop during the early age of the concrete due to thermal gradients and shrinkage effects. Such stresses can cause early-age cracks, termed ‘pre-release cracks’, which occur prior to the transfer of the prestressing force. When the prestressing force is applied to the cross-section, it is assumed that partial or full closure of the cracks occurs by virtue of the force transfer through the cracked cross-section. Verification of the closure of the cracks after the application of the prestressing force is important as it can either confirm continued structural integrity or indicate and approximate reduced structural capacity. Structural health monitoring (SHM) can be used for this purpose. This paper researches an SHM method that can be applied to prestressed beam structures to assess the condition of pre-release cracks. The sensor network used in this method consists of parallel long-gauge fiber optic strain sensors embedded in the concrete cross-sections at various locations. The same network is used for damage detection, i.e. detection and characterization of the pre-release cracks, and for monitoring the prestress force transfer. The method is validated on a real structure, a curved continuous girder. Results from the analysis confirm the safety and integrity of the structure. The method and its application are presented in this paper. (paper)

Analysis of the status of pre-release cracks in prestressed concrete structures using long-gauge sensors

Science.gov (United States)

Abdel-Jaber, H.; Glisic, B.

2015-02-01

Prestressed structures experience limited tensile stresses in concrete, which limits or completely eliminates the occurrence of cracks. However, in some cases, large tensile stresses can develop during the early age of the concrete due to thermal gradients and shrinkage effects. Such stresses can cause early-age cracks, termed ‘pre-release cracks’, which occur prior to the transfer of the prestressing force. When the prestressing force is applied to the cross-section, it is assumed that partial or full closure of the cracks occurs by virtue of the force transfer through the cracked cross-section. Verification of the closure of the cracks after the application of the prestressing force is important as it can either confirm continued structural integrity or indicate and approximate reduced structural capacity. Structural health monitoring (SHM) can be used for this purpose. This paper researches an SHM method that can be applied to prestressed beam structures to assess the condition of pre-release cracks. The sensor network used in this method consists of parallel long-gauge fiber optic strain sensors embedded in the concrete cross-sections at various locations. The same network is used for damage detection, i.e. detection and characterization of the pre-release cracks, and for monitoring the prestress force transfer. The method is validated on a real structure, a curved continuous girder. Results from the analysis confirm the safety and integrity of the structure. The method and its application are presented in this paper.

Life cycle uses of concrete for more sustainable construction

Energy Technology Data Exchange (ETDEWEB)

Horvath, A. [Univ. of California, Berkeley, CA (United States). Dept. of Civil and Environmental Engineering

2001-07-01

This paper examined ways in which the environmental burdens of construction in general and concrete production in particular can be reduced. Aggregates for concrete production include sand, gravel and stone. They account for most (80 per cent) of the materials used in the United States. This paper argued that given the fact that environmental concerns are an important social issue, the issue of natural resource conservation should be addressed. Some of the life-cycle assessments and comparative design issues associated with concrete construction were summarized. The author presented the example that often the initial cost of a new pavement application may indicate a lower environmental impact than an equivalent design when asphalt is used over reinforced concrete. However, annualized impacts may result in comparable environmental assessments. The same is true for bridge girders, reinforced concrete also seems to be a better environmental choice than steel. This paper also described end-of-life options that involve the use of waste products and recycled products in concrete and other materials to reduce the overall environmental impacts of a product or facility. This paper was divided into several sections entitled: life cycle assessments; life cycle inventory assessment of concretes and asphalt pavements; and, life cycle inventory assessment of concrete and steel bridge girders. 16 refs., 4 tabs.

Measurement of lime/silica ratio in concrete using PGNAA technique

Energy Technology Data Exchange (ETDEWEB)

Naqvi, A.A. [Department of Physics, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)]. E-mail: aanaqvi@kfupm.edu.sa; Nagadi, M.M. [Department of Physics, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia); Al-Amoudi, O.S.B. [Department of Civil Engineering, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)

2005-12-01

Prompt gamma neutron activation analysis (PGNAA) technique has been used to determine lime/silica ratio in concrete samples using an accelerator-based PGNAA facility. The ratio was determined from the measured yield of 6.42 MeV prompt {gamma}-rays of calcium and 4.93 MeV {gamma}-rays of silicon from the six concrete samples. The experimental results were compared with the results of the Monte Carlo simulations. An excellent agreement has been achieved between the two. The study has demonstrated successful use of the accelerator-based PGNAA setup in non-destructive analysis of the concrete samples.

Closure Report for Corrective Action Unit 366: Area 11 Plutonium Valley Dispersion Sites, Nevada National Security Site, Nevada

Energy Technology Data Exchange (ETDEWEB)

None, None

2013-12-31

This Closure Report (CR) presents information supporting closure of Corrective Action Unit (CAU) 366, Area 11 Plutonium Valley Dispersion Sites, and provides documentation supporting the completed corrective actions and confirmation that closure objectives for CAU 366 were met. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; the U.S. Department of Energy (DOE), Environmental Management; the U.S. Department of Defense; and DOE, Legacy Management (FFACO, 1996 as amended).

Road Closures

Data.gov (United States)

Montgomery County of Maryland — This is an up to date map of current road closures in Montgomery County.This dataset is updated every few minutes from the Department of Transportation road closure...

The Remote Handled Immobilization Low-Activity Waste Disposal Facility Environmental Permits and Approval Plan

International Nuclear Information System (INIS)

DEFFENBAUGH, M.L.

2000-01-01

The purpose of this document is to revise Document HNF-SD-ENV-EE-003, ''Permitting Plan for the Immobilized Low-Activity Waste Project, which was submitted on September 4, 1997. That plan accounted for the interim storage and disposal of Immobilized-Low Activity Waste at the existing Grout Treatment Facility Vaults (Project W-465) and within a newly constructed facility (Project W-520). Project W-520 was to have contained a combination of concrete vaults and trenches. This document supersedes that plan because of two subsequent items: (1) A disposal authorization that was received on October 25, 1999, in a U. S. Department of Energy-Headquarters, memorandum, ''Disposal Authorization Statement for the Department of Energy Hanford site Low-Level Waste Disposal facilities'' and (2) ''Breakthrough Initiative Immobilized Low-Activity Waste (ILAW) Disposal Alternative,'' August 1999, from Lucas Incorporated, Richland, Washington. The direction within the U. S. Department of Energy-Headquarters memorandum was given as follows: ''The DOE Radioactive Waste Management Order requires that a Disposal authorization statement be obtained prior to construction of new low-level waste disposal facility. Field elements with the existing low-level waste disposal facilities shall obtain a disposal authorization statement in accordance with the schedule in the complex-wide Low-Level Waste Management Program Plan. The disposal authorization statement shall be issued based on a review of the facility's performance assessment and composite analysis or appropriate CERCLA documentation. The disposal authorization shall specify the limits and conditions on construction, design, operations, and closure of the low-level waste facility based on these reviews. A disposal authorization statement is a part of the required radioactive waste management basis for a disposal facility. Failure to obtain a disposal authorization statement or record of decision shall result in shutdown of an operational

Sub-arc narrow gap welding of Atucha 2 RPV closure head

International Nuclear Information System (INIS)

Hantsch, H.; Million, K.; Zimmermann, H.

1982-01-01

Narrow gap technology was used for reasons of design and fabrication when welding the closure-head dome to its flange. Preliminary tests had yielded the necessary improvements of the well-proven sub-arc practice. New facilities had to be developed for welding proper and for the accompanying machining work (finishing in the narrow gap). Special measures were adopted for monitoring the welding process and for recording the welding parameters. The new method was tried out on several large test coupons before welding of the final product was started. No difficulties were encountered during the welding job. Fabrication of the closure head is shown in a short film sequence. (orig.)

Airborne & Field Sensors Facilities

Data.gov (United States)

Federal Laboratory Consortium — RTTC facilities include an 800' x 60' paved UAV operational area, clearapproach/departure zone, concrete pads furnished with 208VAC, 3 phase,200 amp power, 20,000 sq...

Interferometric Imaging Directly with Closure Phases and Closure Amplitudes

Science.gov (United States)

Chael, Andrew A.; Johnson, Michael D.; Bouman, Katherine L.; Blackburn, Lindy L.; Akiyama, Kazunori; Narayan, Ramesh

2018-04-01

Interferometric imaging now achieves angular resolutions as fine as ∼10 μas, probing scales that are inaccessible to single telescopes. Traditional synthesis imaging methods require calibrated visibilities; however, interferometric calibration is challenging, especially at high frequencies. Nevertheless, most studies present only a single image of their data after a process of “self-calibration,” an iterative procedure where the initial image and calibration assumptions can significantly influence the final image. We present a method for efficient interferometric imaging directly using only closure amplitudes and closure phases, which are immune to station-based calibration errors. Closure-only imaging provides results that are as noncommittal as possible and allows for reconstructing an image independently from separate amplitude and phase self-calibration. While closure-only imaging eliminates some image information (e.g., the total image flux density and the image centroid), this information can be recovered through a small number of additional constraints. We demonstrate that closure-only imaging can produce high-fidelity results, even for sparse arrays such as the Event Horizon Telescope, and that the resulting images are independent of the level of systematic amplitude error. We apply closure imaging to VLBA and ALMA data and show that it is capable of matching or exceeding the performance of traditional self-calibration and CLEAN for these data sets.

Restoration activities in uranium mining and milling facilities in Spain

International Nuclear Information System (INIS)

Garcia Quiros, J.M.

1997-01-01

From the end of the 80's up to now, several tasks have been carried out in Spain on restoration in the field of uranium mining and milling, significant among them being Andujar Uranium Mill (FUA) closure and La Haba closure. Also, a study has been carried out on restoration of inoperative and abandoned uranium mine sites. At present, detailed plans are being worked out for the project on the closure of the Elefante plant. All activities have been developed in the common framework of national standards and regulations which are generally in compliance with the standards, regulations and recommendations of international organizations. This paper describes briefly the standards and the criteria applied to the restoration tasks at various sites of the uranium mining and milling facilities in Spain. The restoration activities have different characteristics La Haba facility is an isolated and conventional facility to produce uranium concentrate; in the case of old and abandoned uranium mines the intervention criteria is more relevant than the activities to be carried out; the closure (the first phase of licensing) and restoration activities of Elefante plant have to be developed taking into account that it is sited within the area of Quercus plant which is currently in operation. (author)

Effect of γ-ray irradiation on the properties of concrete

International Nuclear Information System (INIS)

Muraoka, Susumu; Murase, Yoshinobu; Yamada, Kiyotsugu

1983-02-01

In order to obtain the fundamental data of concretes which are used in the construction of storange facilities of HLW, the effect of γ-ray irradiation on the properties of concretes was evaluated. After curing in water for about 4 weeks, concretes were stored under three different condition; 1) heating with γ-ray irradiation, 2) heating without γ-ray irradiation, and 3) controlled at 20 0 C, Relative Humidty of 90%; for 300, 600, and 1200 hours. Then, concretes were tested about the changes of weight, compressive strength, modulus of elasticity, Poisson's ratio, dimension and carbonation depth. Heating temperature were 100, 160 and 190 0 C. Consequently, it was concluded that the effect of γ-ray irradiation within the range of 1.0 x 10 9 R on the properties of concrete was uncertain, and that the effect of heating was more distinguished than γ-ray irradiation effect. (author)

Monitoring of Concrete Structures Using Ofdr Technique

Science.gov (United States)

Henault, J. M.; Salin, J.; Moreau, G.; Delepine-Lesoille, S.; Bertand, J.; Taillade, F.; Quiertant, M.; Benzarti, K.

2011-06-01

Structural health monitoring is a key factor in life cycle management of infrastructures. Truly distributed fiber optic sensors are able to provide relevant information on large structures, such as bridges, dikes, nuclear power plants or nuclear waste disposal facilities. The sensing chain includes an optoelectronic unit and a sensing cable made of one or more optical fibers. A new instrument based on Optical Frequency Domain Reflectometry (OFDR), enables to perform temperature and strain measurements with a centimeter scale spatial resolution over hundred of meters and with a level of precision equal to 1 μstrain and 0.1 °C. Several sensing cables are designed with different materials targeting to last for decades in a concrete aggressive environment and to ensure an optimal transfer of temperature and strain from the concrete matrix to the optical fiber. Tests were carried out by embedding various sensing cables into plain concrete specimens and representative-scale reinforced concrete structural elements. Measurements were performed with an OFDR instrument; meanwhile, mechanical solicitations were imposed to the concrete element. Preliminary experiments are very promising since measurements performed with distributed sensing system are comparable to values obtained with conventional sensors used in civil engineering and with the Strength of Materials Modelling. Moreover, the distributed sensing system makes it possible to detect and localize cracks appearing in concrete during the mechanical loading.

Post-Closure Inspection Report for Corrective Action Unit 90: Area 2 Bitcutter Containment, Nevada Test Site, Nevada

International Nuclear Information System (INIS)

NSTec Environmental Restoration

2007-01-01

Corrective Action Unit (CAU) 90, Area 2 Bitcutter Containment, is identified in the ''Federal Facility Agreement and Consent Order'' of 1996. The post-closure requirements for CAU 90 are described in Section VII.B.8.b of the Nevada Test Site ''Resource Conservation and Recovery Act'' Permit for a Hazardous Waste Management Facility Number NEV HW0021, dated November 2005. Post-closure activities consist of the following: Semiannual inspections of the site using inspection checklists; Photographic documentation; Field note documentation; and Preparation and submittal of an annual Post-Closure Inspection Report. This annual report covers the period of July 2006 to June 2007 and consists of a summary of the results of the inspections, copies of the inspection checklists and field notes, maintenance and repair records (if any), photographs, and conclusions and recommendations. The inspection checklists are provided in Appendix A, a copy of the field notes is provided in Appendix B, and copies of photographs taken during the inspections are provided in Appendix C

Dose and risk assessment of norm Contaminated waste released from trench disposal facility

International Nuclear Information System (INIS)

Abdel Geleel, M.; Ramadan, A.B.; Tawfik, A.A.

2005-01-01

Oil and gas extraction and processing operations accumulate naturally occurring radioactive material (NORM) at concentrations above normal in by-product waste streams. The petroleum industry adopted methods for managing of NORM that are more restrictive than past practices and are likely to provide greater isolation of the radioactivity. Trench was used as a disposal facility for NORM contaminated wastes at one site of the petroleum industry in Egypt. The aim of this work is to calculate the risk and dose assessment received from trench disposal facility directly and after closure (1000 year). RESRAD computer code was used. The results indicated that the total effective dose (TED) received after direct closure of trench disposal facility was 7.7E-4 mSv/y while after 1000 years, it will he 3.4E-4. The health cancer risk after direct closure was 3.3E-8 while after 1000 years post closure it was 6E-8. Results of this assessment will help examine policy issues concerning different options and regulation of NORM contaminated waste generated by petroleum industry

A study on physical properties of concrete and reinforcement at elevated temperatures

International Nuclear Information System (INIS)

Kanazu, Tsutomu

2002-01-01

Reinforced concrete structures such as a containment vessel, a support of the reactor, piping systems and facilities for storing high level radioactive waste in a nuclear power plant are exposed to a high temperature condition. Changes of physical properties of concrete and reinforcement caused by high temperature influence on mechanical behavior of these structures and internal stresses are induced by difference of thermal coefficients between concrete and reinforcement that was reported in the previous paper by the author. These are the special features in high temperature conditions. Temperature dependence of physical properties of concrete and reinforcement are summarized in the paper based on the experimental results. (author)

The impact of hospital closures on geographical access: Evidence from four southeastern states of the United States

Directory of Open Access Journals (Sweden)

M.L. Burkey

Full Text Available This paper examines the effects of hospital closures on geographical access by potential patients, using data from four southeastern U.S. states. Using optimization models designed to minimize the adverse effects of hospital closures, extensive computations are performed and the results are discussed. The effects of the closures on the rural areas is also investigated. Finally, the paper determines which hospitals are most likely among those to be closed assuming that up to 10% of the existing hospitals in each of the four states were to be shut down. The overall conclusion of the empirical findings is that while differences exist among the states, efficiency, coverage, and equality measures for geographical access do not suffer significantly if only a few hospitals are closed in each state, provided these closures are done optimally to minimize impact. Further, for efficiency objectives, decision makers can follow a sequential strategy for closures and still be guaranteed optimality. The paper also discusses the effects of hospital closures on equity and it examines whether or not rural areas are disproportionately affected by closures. Keywords: Health care, Access to health care, Proximity, Hospital closures, Location problems, Facility planning

Development of a low activation concrete shielding wall by multi-layered structure for a fusion reactor

International Nuclear Information System (INIS)

Sato, Satoshi; Maegawa, Toshio; Yoshimatsu, Kenji; Sato, Koichi; Nonaka, Akira; Takakura, Kosuke; Ochiai, Kentaro; Konno, Chikara

2011-01-01

A multi-layered concrete structure has been developed to reduce induced activity in the shielding for neutron generating facilities such as a fusion reactor. The multi-layered concrete structure is composed of: (1) an inner low activation concrete, (2) a boron-doped low activation concrete as the second layer, and (3) ordinary concrete as the outer layer of the neutron shield. With the multi-layered concrete structure the volume of boron is drastically decreased compared to a monolithic boron-doped concrete. A 14 MeV neutron shielding experiment with multi-layered concrete structure mockups was performed at FNS and several reaction rates and induced activity in the mockups were measured. This demonstrated that the multi-layered concrete effectively reduced low energy neutrons and induced activity.

Recycled Concrete as Aggregate for Structural Concrete Production

Directory of Open Access Journals (Sweden)

Mirjana Malešev

2010-04-01

Full Text Available A comparative analysis of the experimental results of the properties of fresh and hardened concrete with different replacement ratios of natural with recycled coarse aggregate is presented in the paper. Recycled aggregate was made by crushing the waste concrete of laboratory test cubes and precast concrete columns. Three types of concrete mixtures were tested: concrete made entirely with natural aggregate (NAC as a control concrete and two types of concrete made with natural fine and recycled coarse aggregate (50% and 100% replacement of coarse recycled aggregate. Ninety-nine specimens were made for the testing of the basic properties of hardened concrete. Load testing of reinforced concrete beams made of the investigated concrete types is also presented in the paper. Regardless of the replacement ratio, recycled aggregate concrete (RAC had a satisfactory performance, which did not differ significantly from the performance of control concrete in this experimental research. However, for this to be fulfilled, it is necessary to use quality recycled concrete coarse aggregate and to follow the specific rules for design and production of this new concrete type.

How Concrete Is Concrete?

Science.gov (United States)

Gravemeijer, Koeno

2011-01-01

If we want to make something concrete in mathematics education, we are inclined introduce, what we call, "manipulatives", in the form of tactile objects or visual representations. If we want to make something concrete in a everyday-life conversation, we look for an example. In the former, we try to make a concrete model of our own,…

Influence of Different Container Closure Systems and Capping Process Parameters on Product Quality and Container Closure Integrity (CCI) in GMP Drug Product Manufacturing.

Science.gov (United States)

Mathaes, Roman; Mahler, Hanns-Christian; Roggo, Yves; Huwyler, Joerg; Eder, Juergen; Fritsch, Kamila; Posset, Tobias; Mohl, Silke; Streubel, Alexander

2016-01-01

Capping equipment used in good manufacturing practice manufacturing features different designs and a variety of adjustable process parameters. The overall capping result is a complex interplay of the different capping process parameters and is insufficiently described in literature. It remains poorly studied how the different capping equipment designs and capping equipment process parameters (e.g., pre-compression force, capping plate height, turntable rotating speed) contribute to the final residual seal force of a sealed container closure system and its relation to container closure integrity and other drug product quality parameters. Stopper compression measured by computer tomography correlated to residual seal force measurements.In our studies, we used different container closure system configurations from different good manufacturing practice drug product fill & finish facilities to investigate the influence of differences in primary packaging, that is, vial size and rubber stopper design on the capping process and the capped drug product. In addition, we compared two large-scale good manufacturing practice manufacturing capping equipment and different capping equipment settings and their impact on product quality and integrity, as determined by residual seal force.The capping plate to plunger distance had a major influence on the obtained residual seal force values of a sealed vial, whereas the capping pre-compression force and the turntable rotation speed showed only a minor influence on the residual seal force of a sealed vial. Capping process parameters could not easily be transferred from capping equipment of different manufacturers. However, the residual seal force tester did provide a valuable tool to compare capping performance of different capping equipment. No vial showed any leakage greater than 10(-8)mbar L/s as measured by a helium mass spectrometry system, suggesting that container closure integrity was warranted in the residual seal force range

Usage of Crushed Concrete Fines in Decorative Concrete

Science.gov (United States)

Pilipenko, Anton; Bazhenova, Sofia

2017-10-01

The article is devoted to the questions of usage of crushed concrete fines from concrete scrap for the production of high-quality decorative composite materials based on mixed binder. The main problem in the application of crushed concrete in the manufacture of decorative concrete products is extremely low decorative properties of crushed concrete fines itself, as well as concrete products based on them. However, crushed concrete fines could have a positive impact on the structure of the concrete matrix and could improve the environmental and economic characteristics of the concrete products. Dust fraction of crushed concrete fines contains non-hydrated cement grains, which can be opened in screening process due to the low strength of the contact zone between the hydrated and non-hydrated cement. In addition, the screening process could increase activity of the crushed concrete fines, so it can be used as a fine aggregate and filler for concrete mixes. Previous studies have shown that the effect of the usage of the crushed concrete fines is small and does not allow to obtain concrete products with high strength. However, it is possible to improve the efficiency of the crushed concrete fines as a filler due to the complex of measures prior to mixing. Such measures may include a preliminary mechanochemical activation of the binder (cement binder, iron oxide pigment, silica fume and crushed concrete fines), as well as the usage of polycarboxylate superplasticizers. The development of specific surface area of activated crushed concrete fines ensures strong adhesion between grains of binder and filler during the formation of cement stone matrix. The particle size distribution of the crushed concrete fines could achieve the densest structure of cement stone matrix and improve its resistance to environmental effects. The authors examined the mechanisms of structure of concrete products with crushed concrete fines as a filler. The results of studies of the properties of

High Performance Zero-Bleed CLSM/Grout Mixes for High-Level Waste Tank Closures Strategic Research and Development - FY99 Report

International Nuclear Information System (INIS)

Langton, C.A.

2000-01-01

The overall objective of this program, SRD-99-08, was to design and test suitable materials, which can be used to close high-level waste tanks at SRS. Fill materials can be designed to perform several functions including chemical stabilization and/or physical encapsulation of incidental waste so that the potential for transport of contaminants into the environment is reduced. Also they are needed to physically stabilize the void volume in the tanks to prevent/minimize future subsidence and inadvertent intrusion. The intent of this work was to develop a zero-bleed soil CLSM (ZBS-CLSM) and a zero-bleed concrete mix (ZBC) which meet the unique placement and stabilization/encapsulation requirements for high-level waste tank closures. These mixes in addition to the zero-bleed CLSM mixes formulated for closure of Tanks 17-F and 20-F provide design engineers with a suite of options for specifying materials for future tank closures

In-Situ Testing and Performance Assessment of a Redesigned WIPP Panel Closure - 13192

Energy Technology Data Exchange (ETDEWEB)

Klein, Thomas [URS-Professional Solutions, 4021 National Parks Highway Carlsbad, NM 88220 (United States); Patterson, Russell [Department of Energy-Carlsbad Field Office, 4021 National Parks Highway Carlsbad, NM 88220 (United States); Camphouse, Chris; Herrick, Courtney; Kirchner, Thomas; Malama, Bwalya; Zeitler, Todd [Sandia National Laboratories-Carlsbad, 4100 National Parks Highway Carlsbad, NM 88220 (United States); Kicker, Dwayne [SM Stoller Corporation-Carlsbad, 4100 National Parks Highway Carlsbad, NM (United States)

2013-07-01

There are two primary regulatory requirements for Panel Closures at the Waste Isolation Pilot Plant (WIPP), the nation's only deep geologic repository for defense related Transuranic (TRU) and Mixed TRU waste. The Federal requirement is through 40 CFR 191 and 194, promulgated by the U.S. Environmental Protection Agency (EPA). The state requirement is regulated through the authority of the Secretary of the New Mexico Environment Department (NMED) under the New Mexico Hazardous Waste Act (HWA), New Mexico Statutes Annotated (NMSA) 1978, chap. 74-4-1 through 74-4-14, in accordance with the New Mexico Hazardous Waste Management Regulations (HWMR), 20.4.1 New Mexico Annotated Code (NMAC). The state regulations are implemented for the operational period of waste emplacement plus 30 years whereas the federal requirements are implemented from the operational period through 10,000 years. The 10,000 year federal requirement is related to the adequate representation of the panel closures in determining long-term performance of the repository. In Condition 1 of the Final Certification Rulemaking for 40 CFR Part 194, the EPA required a specific design for the panel closure system. The U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) has requested, through the Planned Change Request (PCR) process, that the EPA modify Condition 1 via its rulemaking process. The DOE has also requested, through the Permit Modification Request (PMR) process, that the NMED modify the approved panel closure system specified in Permit Attachment G1. The WIPP facility is carved out of a bedded salt formation 655 meters below the surface of southeast New Mexico. Condition 1 of the Final Certification Rulemaking specifies that the waste panels be closed using Option D which is a combination of a Salado mass concrete (SMC) monolith and an isolation/explosion block wall. The Option D design was also accepted as the panel closure of choice by the NMED. After twelve years of waste handling

Concrete mix design for X-and gamma shielding

International Nuclear Information System (INIS)

Mohamad Pauzi Ismail; Noor Azreen Masenwat; Suhairy Sani; Abdul Bakhri Muhammad; Mohd Kamal Shah Shamsuddin; Rahmad Abd Rashid

2012-01-01

The design of X-ray or gamma ray radiographic exposure room requires some calculations on shielding to provide safe operation of the facility and minimum exposure to radiation workers. Careful design can lead to economical installations with minimal barriers. The design depends on such factors as: maximum energy, maximum intensity, permitted full-body dosage, workload, use factor, occupancy factor, maximum dose output and shielding materials. Choice of material for a barrier depends on convenience and cost. The radiographic exposure room is usually made of normal concrete with density of about 2.3 - 2.4 g/ cc. Normal concrete is often used for construction of exposure room because of cheap and ease of construction. This paper explained and discussed the optimum mix design for normal concrete used for X-and gamma shielding. (author)

Corrective Action Plan for CAU No. 95: Area 15 EPA Farm Laboratory Building, Decontamination and Demolition Closure Activities - Nevada Test Site. Rev. 0

International Nuclear Information System (INIS)

Olson, A.L.; Nacht, S.J.

1997-11-01

This Corrective Action Plan (CAP) provides the selected corrective action alternative and proposes the closure implementation methodology for the Environmental Protection Agency (EPA) Farm Laboratory Building 15-06 located in Area 15 of the Nevada Test Site (NTS), Nye County, Nevada. The facility is part of the Environmental Restoration Project managed by the U.S. Department of Energy/Nevada Operations Office (DOE/NV) under the Decontamination and Decommissioning (D ampersand D) Subproject which serves to manage and dispose of surplus facilities at the NTS in a manner that will protect personnel, the public, and the environment. It is identified as Corrective Action Unit (CAU) 95 in Appendix III of the Federal Facilities Agreement and Consent Order (FFACO). In July 1997, the DOE/NV verbally requested approval from the Nevada Division of Environmental Protection (NDEP) for the closure schedule to be accelerated. Currently, field activities are anticipated to be completed by September 30, 1997. In order to meet this new schedule NDEP has agreed to review this document as expeditiously as possible. Comments will be addressed in the Closure Report after field activities have been completed, unless significant issues require resolution during closure activities

Corrective Action Plan for CAU No. 95: Area 15 EPA Farm Laboratory Building, Decontamination and Demolition Closure Activities - Nevada Test Site. Rev. 0

Energy Technology Data Exchange (ETDEWEB)

Olson, A.L.; Nacht, S.J.

1997-11-01

This Corrective Action Plan (CAP) provides the selected corrective action alternative and proposes the closure implementation methodology for the Environmental Protection Agency (EPA) Farm Laboratory Building 15-06 located in Area 15 of the Nevada Test Site (NTS), Nye County, Nevada. The facility is part of the Environmental Restoration Project managed by the U.S. Department of Energy/Nevada Operations Office (DOE/NV) under the Decontamination and Decommissioning (D&D) Subproject which serves to manage and dispose of surplus facilities at the NTS in a manner that will protect personnel, the public, and the environment. It is identified as Corrective Action Unit (CAU) 95 in Appendix III of the Federal Facilities Agreement and Consent Order (FFACO). In July 1997, the DOE/NV verbally requested approval from the Nevada Division of Environmental Protection (NDEP) for the closure schedule to be accelerated. Currently, field activities are anticipated to be completed by September 30, 1997. In order to meet this new schedule NDEP has agreed to review this document as expeditiously as possible. Comments will be addressed in the Closure Report after field activities have been completed, unless significant issues require resolution during closure activities.

Closure The Definitive Guide

CERN Document Server

Bolin, Michael

2010-01-01

If you're ready to use Closure to build rich web applications with JavaScript, this hands-on guide has precisely what you need to learn this suite of tools in depth. Closure makes it easy for experienced JavaScript developers to write and maintain large and complex codebases -- as Google has demonstrated by using Closure with Gmail, Google Docs, and Google Maps. Author and Closure contributor Michael Bolin has included numerous code examples and best practices, as well as valuable information not available publicly until now. You'll learn all about Closure's Library, Compiler, Templates, tes

Thermal studies of the canister staging pit in a hypothetical Yucca Mountain canister handling facility using computational fluid dynamics

International Nuclear Information System (INIS)

Soltani, Mehdi; Barringer, Chris; Bues, Timothy T. de

2007-01-01

The proposed Yucca Mountain nuclear waste storage site will contain facilities for preparing the radioactive waste canisters for burial. A previous facility design considered was the Canister Handling Facility Staging Pit. This design is no longer used, but its thermal evaluation is typical of such facilities. Structural concrete can be adversely affected by the heat from radioactive decay. Consequently, facilities must have heating ventilation and air conditioning (HVAC) systems for cooling. Concrete temperatures are a function of conductive, convective and radiative heat transfer. The prediction of concrete temperatures under such complex conditions can only be adequately handled by computational fluid dynamics (CFD). The objective of the CFD analysis was to predict concrete temperatures under normal and off-normal conditions. Normal operation assumed steady state conditions with constant HVAC flow and temperatures. However, off-normal operation was an unsteady scenario which assumed a total HVAC failure for a period of 30 days. This scenario was particularly complex in that the concrete temperatures would gradually rise, and air flows would be buoyancy driven. The CFD analysis concluded that concrete wall temperatures would be at or below the maximum temperature limits in both the normal and off-normal scenarios. While this analysis was specific to a facility design that is no longer used, it demonstrates that such facilities are reasonably expected to have satisfactory thermal performance. (author)

Thermal stress analysis of the fuel storage facility

International Nuclear Information System (INIS)

Chen, W.W.

1991-12-01

This paper presents the results of a nonlinear finite-element analysis to determine the structural integrity of the walls of the nuclear fuel storage room in the Radio Isotope Power System Facility of the Fuels and Materials Examination Facility (FMEF) Project. The analysis was performed to assess the effects of thermal loading on the walls that would result from a loss-of-cooling accident. The results obtained from using the same three-dimensional finite-element model with different types of elements, the eight-node brick element and the nonlinear concrete element, and the calculated results using the analytical solutions, are compared. The concrete responses in terms of octahedral normal and shearing stresses are described. The crack and crush states of the concrete were determined on the basis of multiaxial failure criteria

Recycled Concrete as Aggregate for Structural Concrete Production

OpenAIRE

Mirjana Malešev; Vlastimir Radonjanin; Snežana Marinković

2010-01-01

A comparative analysis of the experimental results of the properties of fresh and hardened concrete with different replacement ratios of natural with recycled coarse aggregate is presented in the paper. Recycled aggregate was made by crushing the waste concrete of laboratory test cubes and precast concrete columns. Three types of concrete mixtures were tested: concrete made entirely with natural aggregate (NAC) as a control concrete and two types of concrete made with natural fine and recycle...

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.

Site characterization of ORNL D ampersand D facilities

International Nuclear Information System (INIS)

Kelsey, A.P.; Mandry, G.J.; Haghighi, M.H.

1994-01-01

Site characterization for decontamination and decommissioning (D ampersand D) planning purposes was done for two surplus facilities at Oak Ridge National Laboratory (ORNL) in late 1993 and early 1994. This site characterization includes measurements of radiological and chemical contaminants, assessment of general structural conditions, and investigation of unknown conditions within the buildings. It will serve as input to decisions on D ampersand D engineering, D ampersand D task sequences, radiological and contamination control, and waste management. This paper presents the methods used to investigate these facilities and discusses the preliminary results as they apply to D ampersand D planning. Investigation methods include gross alpha, beta, and gamma surveys; directional gamma surveys; gamma spectroscopy; concrete coring; photography; and collection of soil and miscellaneous samples that are analyzed for radiological and chemical contaminants. Data will be analyzed using radiological models to sort sources and estimate exposure rates and waste volumes due to D ampersand D. The former Waste Evaporator Facility (WEF), consisting of two concrete cells and an operating gallery, once contained a liquid radwaste evaporator. Subsequently it was used for an incinerator experiment and as a dressing area for remediation work on an adjacent tank farm. The building has been partially decontaminated. Figure 1 is a photograph of the WEF. The Fission Product Pilot Plant (FPPP) is a small concrete building containing two cells. It was used to extract isotopes of ruthenium, strontium, cesium, cerium, and other elements from liquid waste. This facility is highly contaminated. In 1960 all doors into FPPP were sealed with concrete block and mortar, and concrete block shielding was added to the external walls making them up to five feet thick. Prior to this study, almost nothing was known about the interior of this building

Review of Research on and Implementation of Recycled Concrete Aggregate in the GCC

Directory of Open Access Journals (Sweden)

Akmal S. Abdelfatah

2011-01-01

Full Text Available The goal of sustainable construction is to reduce the environmental impact of a constructed facility over its lifetime. Concrete is the main material used in construction in the Gulf Cooperation Council (GCC. Therefore, it makes economic and environmental sense to use recycled materials in the making of new concrete for different applications. The objectives of this study are to summarize published research on the use of recycled concrete aggregates in new concrete mixes and examine its implementation in construction and industry in the GCC region. The study showed that while there is reasonable research on recycled concrete, the practical implementation in the region greatly lacks behind, especially due to the lack of economic viability and awareness of such applications at the current time.

The Future Concrete: Self-Compacting Concrete

OpenAIRE

Iureş, Liana; Bob, Corneliu

2010-01-01

The paper presents the characteristics of the self-compacting concretes, their advantages and disadvantages when they are used in buildings. Due to its properties and composition, the self-compacting concrete is described here as being one of the future friendly enviromental material for buildings. Tests concerning to obtaining a self-compacting concrete, together with the specific fresh concrete properties tests, are described.

Determining prestressing forces for inspection of prestressed concrete containments

International Nuclear Information System (INIS)

1990-07-01

General Design Criterion 53, ''Provisions for Containment Testing and Inspection,'' of Appendix A, ''General Design Criteria for Nuclear Power Plants,'' to 10 CFR Part 50, ''Domestic Licensing of Production and Utilization Facilities,'' requires, in part, that the reactor containment be designed to permit (1) periodic inspection of all important areas and (2) an appropriate surveillance program. Regulatory Guide 1.35, ''Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures,'' describes a basis acceptable to the NRC staff for developing an appropriate inservice inspection and surveillance program for ungrouted tendons in prestressed concrete containment structures of light-water-cooled reactors. This guide expands and clarifies the NRC staff position on determining prestressing forces to be used for inservice inspections of prestressed concrete containment structures

Centrifugal shot blasting. Innovative technology summary report

International Nuclear Information System (INIS)

1999-07-01

At the US Department of Energy (DOE) Fernald Environmental Management Project (FEMP), the Facilities Closure and Demolition Projects Integrated Remedial Design/Remedial Action (RD/RA) work plan calls for the removal of one inch (1 in) depth of concrete surface in areas where contamination with technetium-99 has been identified. This report describes a comparative demonstration between two concrete removal technologies: an innovative system using Centrifugal Shot Blasting (CSB) and a modified baseline technology called a rotary drum planer

Abrasive-waterjet cutting of thick concrete and waterjet cleaning for nuclear facility decommissioning and decontamination

International Nuclear Information System (INIS)

Echert, D.C.; Hashish, M.; Marvin, M.H.

1987-01-01

Two tools have been developed for use by the nuclear industry: the Deep Kerf tool and the Cleaner/Scarifier tool. The Deep Kerf tool is designed to cut through thick, reinforced concrete structures to facilitate their decommissioning. It employs the abrasive-waterjet (AWJ) cutting technology. The basis of the system is a rotary nozzle that makes a slot in the concrete wide enough to accommodate the cutting tool as it advances. In this program, concrete as thick as 1.5 m was cut through from one side. A shroud and vacuum system covers the opening of the slot during cutting to contain the spoils with greater than 99% efficiency. The Cleaner/Scarifier tool was designed for removing the surface layers of contaminated concrete and decontaminating metal surfaces. It uses ultrahigh-pressure waterjets mounted on a rotating arm to remove or clean the target surface. Spoils recovery with a shroud and vacuum system is over 99% complete for both horizontal and vertical surfaces

Closure Report Central Nevada Test Area Subsurface Corrective Action Unit 443 January 2016

Energy Technology Data Exchange (ETDEWEB)

Findlay, Rick [US Department of Energy, Washington, DC (United States). Office of Legacy Management

2015-11-01

The U.S. Department of Energy (DOE) Office of Legacy Management (LM) prepared this Closure Report for the subsurface Corrective Action Unit (CAU) 443 at the Central Nevada Test Area (CNTA), Nevada, Site. CNTA was the site of a 0.2- to 1-megaton underground nuclear test in 1968. Responsibility for the site’s environmental restoration was transferred from the DOE, National Nuclear Security Administration, Nevada Field Office to LM on October 1, 2006. The environmental restoration process and corrective action strategy for CAU 443 are conducted in accordance with the Federal Facility Agreement and Consent Order (FFACO 1996, as amended 2011) and all applicable Nevada Division of Environmental Protection (NDEP) policies and regulations. This Closure Report provides justification for closure of CAU 443 and provides a summary of completed closure activities; describes the selected corrective action alternative; provides an implementation plan for long-term monitoring with well network maintenance and approaches/policies for institutional controls (ICs); and presents the contaminant, compliance, and use-restriction boundaries for the site.

Experimental study on the leakage of gas through cracked concrete walls

International Nuclear Information System (INIS)

Suzuki, T.; Takiguchi, K.; Hotta, H.; Kojima, N.; Fukuhara, M.; Kimura, K.

1989-01-01

The air-tightness of concrete walls is important for nuclear-related facilities. A concrete wall has very high probabilities of developing cracks due to shrinkage, seismic forces or other factors. It is therefore essential to be able to predict the amount of gas which will leak through a cracked concrete wall. In the previous paper published in SmiRT-9, the experimental equation on the gas leakage through a single crack occurred in concrete was presented based on two-dimensional Poiseuille's flow. In this paper, the experimental results were examined again considering the compressibility of gas, and new equation is presented. The experiments which were similar to ones in the previous paper were carried out on several kinds of concrete using several kinds of gases, and the effects of the kinds of gaseous body, particle size of aggregates and shape of aggregates were examined

Health physics challenges involved with opening a "seventeen-inch" concrete waste vault.

Science.gov (United States)

Sullivan, Patrick T; Pizzulli, Michelle

2005-05-01

This paper describes the various activities involved with opening a sealed legacy "Seventeen-inch" concrete vault and the health physics challenges and solutions employed. As part of a legacy waste stream that was removed from the former Hazardous Waste Management Facility at Brookhaven National Laboratory, the "Seventeen-inch" concrete vault labeled 1-95 was moved to the new Waste Management Facility for ultimate disposal. Because the vault contained 239Pu foils with a total activity in excess of the transuranic waste limits, the foils needed to be removed and repackaged for disposal. Conventional diamond wire saws could not be used because of facility constraints, so this project relied mainly on manual techniques. The planning and engineering controls put in place enabled personnel to open the vault and remove the waste while keeping dose as low as reasonably achievable.

A new concept for the targeted cutting of concrete structures

International Nuclear Information System (INIS)

Reinhardt, Steffen; Gentes, Sascha; Weidemann, Roman; Geimer, Marcus

2011-01-01

The decontamination and crushing of reinforced concrete is a main part during deconstruction of nuclear facilities. The selective treatment of contaminated or activated material is of special interest, since the non-contaminated material can be transferred into the normal reprocessing cycle. In the frame of a project concerning the innovative cutting of massive reinforced concrete structures an all-purpose system for spatially restricted and defined cutting of strongly reinforced concrete including packaging suitable for final disposal was developed. Due to the remote handling of the machine the dose rate for personnel can be reduced significantly. Main part of the system is the tool that can cut highly reinforced concrete without system or component replacement. The authors describe preliminary tests of these tools, further experiments and process optimization are necessary before the tools can be integrated into the new system.

Impact of Different Standard Type A7A Drum Closure-Ring Practices on Gasket Contraction and Bolt Closure Distance– 15621

Energy Technology Data Exchange (ETDEWEB)

Ketusky, Edward [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Blanton, Paul [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Bobbitt, John H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-03-11

The Department of Energy, the Savannah River National Laboratory, several manufacturers of specification drums, and the United States Department of Transportation (DOT) are collaborating in the development of a guidance document for DOE contractors and vendors who wish to qualify containers to DOT 7A Type A requirements. Currently, the effort is focused on DOT 7A Type A 208-liter (55-gallons) drums with a standard 12-gauge bolted closure ring. The U.S. requirements, contained in Title 49, Part 178.350 “Specification 7A; general packaging, Type A specifies a competent authority review of the packaging is not required for the transport of (Class 7) radioactive material containing less than Type A quantities of radioactive material. For Type AF drums, a 4 ft. regulatory free drop must be performed, such that the drum “suffers maximum damage.” Although the actual orientation is not defined by the specification, recent studies suggest that maximum damage would result from a shallow angle top impact, where kinetic energy is transferred to the lid, ultimately causing heavy damage to the lid, or even worse, causing the lid to come off. Since each vendor develops closure recommendations/procedures for the drums they manufacture, key parameters applied to drums during closing vary based on vendor. As part of the initial phase of the collaboration, the impact of the closure variants on the ability of the drum to suffer maximum damage is investigated. Specifically, closure testing is performed varying: 1) the amount of torque applied to the closure ring bolt; and, 2) stress relief protocol, including: a) weight of hammer; and, b) orientation that the hammer hits the closure ring. After closure, the amount of drum lid gasket contraction and the distance that the closure bolt moves through the closure ring is measured.

Current status of JAERI Tokai hot cell facilities

International Nuclear Information System (INIS)

Itami, Hiroharu; Morozumi, Minoru; Yamahara, Takeshi

1992-01-01

JAERI has 4 hot cell facilities in order to examine high radioactive materials. Three of them, the Research Hot Laboratory, the Reactor Fuel Examination Facility and the Waste Safety Testing Facility are located in the JAERI Tokai site, and the rest is the JMTR Hot Laboratory in the Oarai site. The Research Hot Laboratory (RHL) was constructed for post-irradiation examination (PIE), especially nuclear related basic research experiment, such as metallurgical, chemical and mechanical examination on fuels and materials irradiated in research and test reactors. This facility has 10 large dimension concrete and 38 lead cells. At present the RHL is used for various kinds of examinations of high radioactive samples such as fuels of research and test reactors, power reactors and high temperature testing reactor (HTTR), and structural materials. The Reactor Fuel Examination Facility (RFEF) was designed and constructed for carrying out PIE of irradiated full-size fuel assemblies of light water reactors (LWRs). This facility has a storage pool, 8 concrete and 5 lead cells. They are currently used for safety evaluation on high burnup and advanced lWR fuels as part of the national program. The Waste Safety Testing Facility (WASTEF) was designed and constructed for safety research on long-term storage and disposal of high level radioactive wastes, generated by fuel reprocessing. The WASTEF has 5 concrete cells and 1 lead cell. Examinations on the behavior of various long-lived fission products in a glass form and in a canister and, releasing behavior of them out of a canister are carrying out under the condition at storage. (author)

The Future Concrete: Self-Compacting Concrete

Directory of Open Access Journals (Sweden)

Liana Iureş

2010-01-01

Full Text Available The paper presents the characteristics of the self-compacting concretes, their advantages and disadvantages when they are used in buildings. Due to its properties and composition, the self-compacting concrete is described here as being one of the future friendly enviromental material for buildings. Tests concerning to obtaining a self-compacting concrete, together with the specific fresh concrete properties tests, are described.

Forterra Concrete Products, Inc. - Clean Water Act Public Notice

Science.gov (United States)

The EPA is providing notice of a proposed Administrative Penalty Assessment against Forterra Concrete Products, Inc., a business located at 511 E. John Carpenter Freeway, Irving, TX, 75062, for alleged violations at its facility located at 23600 W. 40th St

Offshore concrete structures; Estructuras Offshore (mar adentro) de Hormigon

Energy Technology Data Exchange (ETDEWEB)

Lamas Pardo, M.; Carral Couce, L. M.

2011-07-01

proof that in certain cases the benefits to the steel tip the lance toward the concrete. concrete Gravity Baseplaforms type Con deep have been building since the first unit became operational in 1976, together with a few floating platforms, that in its geometry are similar to those made of steel (barges, semi-submersibles and TLP type platforms). Some of the concepts in concrete that are being emerging last years are floating and gravity base platforms for use as offshore LNG terminals and as offshore industrial plants. The life of these barges can be designed up to 200 years, as the floating Nkossabarge, so they con be a good alternative to the construction of these facilities an land, thus avoiding landfills on the coast that degrade the already punished coast in industrialized countries. the challenge is precisely to optimize their capital costs to complete for an offshore installation against a shore facility. The environment will undoubtedly benefit from this great challenge that lies ahead in the XXI century. (Author) 29 refs.

Activation of air and concrete in medical isotope production facilities

Science.gov (United States)

Dodd, Adam C.; Shackelton, R. J.; Carr, D. A.; Ismail, A.

2017-05-01

Medical isotope facilities operating in the 10 to 25 MeV proton energy range have long been used to generate radioisotopes for medical diagnostic imaging. In the last few years the beam currents available in commercially available cyclotrons have increased dramatically, and so the activation of the materials within cyclotron vaults may now pose more serious radiological hazards. This will impact the regulatory oversight of cyclotron operations, cyclotron servicing and future decommissioning activities. Air activation could pose a hazard to cyclotron staff. With the increased cyclotron beam currents it was necessary to examine the issue more carefully. Therefore the ways in which radioactivity may be induced in air by neutron reactions and neutron captures were considered and it was found that the dominant mechanism is neutron capture on Ar-40. A study of the activation of the air by neutron capture on Ar-40 within a cyclotron vault was performed using the MCNP Monte Carlo code. The neutron source energy spectrum used was from the production of the widely used F-18 PET isotope. The results showed that the activation of the air within a cyclotron vault does not pose a significant radiological hazard at the beam intensities currently in use and shows how ventilation affects the results. A second MCNP study on the activation of ordinary concrete in cyclotron vaults by neutron capture was made with a view to determining the optimum thickness of borated polyethylene to reduce neutron activation on both the inner surfaces of the vault and around production targets. This is of importance in decommissioning cyclotrons and therefore in the design of new cyclotron vaults. The distribution of activation on the walls as a function of the source position was also studied. Results are presented for both borated and regular polyethylene, and F-18 and Tc-99 neutron spectra.

Concrete under severe conditions. Environment and loading

International Nuclear Information System (INIS)

2007-01-01

-disciplinary ideas in order to improve the current situation. In continuity with previous CONSEC conferences, chloride ingress within concrete structures, durability performance in marine environment, and mitigation of corrosion has deserved important research effort, leading to numerous meaningful contributions. Themes of wear, fatigue resistance of concrete and concrete structures, and control of crack openings represent traditionally important durability issues, for which significant new results are provided. Moreover, with the evolution of research themes, of design and building practice and of social needs, some topics indicate a renewed and major interest of CONSEC'07 contributors: - shock and impact strength of concrete structures, in relation with natural and industrial hazards or terrorist attacks, - thermo-mechanical couplings and fire resistance of concrete structures, which represents an important industrial and safety issue, - rational accounting for chemical attacks and prevention of associated degradations, - performance-based engineering of concrete material, with a special care of durability. Sixteen papers dealing with the behaviour of concrete structures in nuclear facilities have been selected for INIS

Closure plan for the M-Area settling basin and vicinity at the Savannah River Plant

International Nuclear Information System (INIS)

Colven, W.P.; Pickett, J.B.; Muska, C.F.; Boone, L.F.

1988-03-01

The closure plan for the M-Area settling basin and vicinity was originally submitted to the South Carolina Department of Health and Environmental Control (SCDHEC) Bureau of Solid and Hazardous Waste Management in September 1984. The plan was revised in July and November 1985 in response to SCDHEC comments. After public comment in April through July 1986, the closure plan was conditionally approved by SCDHEC in March 1987. The conditions included (1) providing a temporary wastewater treatment facility to process the water remaining in the basin, (2) using a burn box to limit ash emissions from burning in the basin, (3) obtaining SCDHEC approval prior to operating the leach field, and (4) completing all closure activities within three years of the startup date

NPDES Permit for Super Concrete Ready-Mix Corp. (Aggregate Industries)

Science.gov (United States)

Under National Pollutant Discharge Elimination System permit number DC0000175, Super Concrete Ready-Mix Corporation is authorized to discharge from a facility to receiving waters named unnamed tributary to Northwest Branch of the Anacostia River.

Elevated Fixed Platform Test Facility

Data.gov (United States)

Federal Laboratory Consortium — The Elevated Fixed Platform (EFP) is a helicopter recovery test facility located at Lakehurst, NJ. It consists of a 60 by 85 foot steel and concrete deck built atop...

Radionuclide Retention in Concrete Wasteforms - FY13

Energy Technology Data Exchange (ETDEWEB)

Snyder, Michelle MV; Golovich, Elizabeth C.; Wellman, Dawn M.; Crum, Jarrod V.; Lapierre, Robert; Dage, Denomy C.; Parker, Kent E.; Cordova, Elsa A.

2013-10-15

Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation); the mechanism of contaminant release; the significance of contaminant release pathways; how wasteform performance is affected by the full range of environmental conditions within the disposal facility; the process of wasteform aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of wasteform aging on chemical, physical, and radiological properties; and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the wasteforms come in contact with groundwater. Data collected throughout the course of this work will be used to quantify the efficacy of concrete wasteforms, similar to those used in the disposal of low-level waste and mixed low-level waste, for the immobilization of key radionuclides (i.e., uranium, technetium, and iodine). Data collected will also be used to quantify the physical and chemical properties of the concrete affecting radionuclide retention.

Supporting Fernald Site Closure with Integrated Health and Safety Plans as Documented Safety Analyses

International Nuclear Information System (INIS)

Kohler, S.; Brown, T.; Fisk, P.; Krach, F.; Klein, B.

2004-01-01

At the Fernald Closure Project (FCP) near Cincinnati, Ohio, environmental restoration activities are supported by Documented Safety Analyses (DSAs) that combine the required project-specific Health and Safety Plans, Safety Basis Requirements (SBRs), and Process Requirements (PRs) into single Integrated Health and Safety Plans (I-HASPs). These integrated DSAs employ Integrated Safety Management methodology in support of simplified restoration and remediation activities that, so far, have resulted in the decontamination and demolition (D and D) of over 200 structures, including eight major nuclear production plants. There is one of twelve nuclear facilities still remaining (Silos containing uranium ore residues) with its own safety basis documentation. This paper presents the status of the FCP's safety basis documentation program, illustrating that all of the former nuclear facilities and activities have now replaced. Basis of Interim Operations (BIOs) with I-HASPs as their safety basis during the closure process

Post-Closure Inspection Report for Corrective Action Unit 90: Area 2 Bitcutter Containment, Nevada Test Site, Nevada

International Nuclear Information System (INIS)

NSTec Environmental Restoration

2008-01-01

Corrective Action Unit (CAU) 90, Area 2 Bitcutter Containment, is identified in the Federal Facility Agreement and Consent Order of 1996, as amended February 2008. The post-closure requirements for CAU 90 are described in Section VII.B.8.b of the Nevada Test Site Resource Conservation and Recovery Act Permit for a Hazardous Waste Management Facility Number NEV HW0021, dated November 2005. Post-closure activities consist of the following: (1) Semiannual inspections of the site using inspection checklists; (2) Photographic documentation; (3) Field note documentation; and (4) Preparation and submittal of an annual Post-Closure Inspection Report. This annual report covers the period of July 2007 to June 2008 and consists of a summary of the results of the inspections, copies of the inspection checklists and field notes, maintenance and repair records (if any), photographs, and conclusions and recommendations. The inspection checklists are provided in Appendix A, a copy of the field notes is provided in Appendix B, and copies of photographs taken during the inspections are provided in Appendix C

Professional Closure Beyond State Authorization

Directory of Open Access Journals (Sweden)

Gitte Sommer Harrits

2014-03-01

Full Text Available For decades, the Weberian approach to the study of professions has been strong, emphasizing state authorization and market monopolies as constituting what is considered a profession. Originally, however, the Weberian conception of closure, or the ways in which a profession is constituted and made separate, was broader. This article suggests a revision of the closure concept, integrating insights from Pierre Bourdieu, and conceptualizing professional closure as the intersection of social, symbolic and legal closure. Based on this revision, this article demonstrates how to apply such a concept in empirical studies. This is done by exploring social, symbolic and legal closure across sixteen professional degree programs. The analyses show a tendency for some overlap between different forms of closure, with a somewhat divergent pattern for legal closure. Results support the argument that we need to study these processes as an intersection of different sources of closure, including capital, lifestyles and discourse

Clearance of concrete debris generated from modification work of JRR-3

International Nuclear Information System (INIS)

Satoyama, Tomonori; Nanri, Tomohiro; Kishimoto, Katsumi

2014-01-01

The Japan Atomic Energy Agency (JAEA) planned to apply the clearance system to slightly contaminated concrete debris with radionuclides, which was generated from the modification of the Japan Research Reactor No. 3 (JRR-3) in the Nuclear Science Research Institute (NSRI) of JAEA. The modification work was conducted from 1985 to 1990 and the generated concrete debris has been stored as radioactive waste in interim storage facilities in the NSRI. This is the Japanese first approved of clearance system to concrete debris and stored waste. We established procedures for measuring and evaluating the radioactivity concentration of scored concrete debris. In 2008, the authority approved these procedures for clearance. Since 2009, we have been measuring and evaluating the radioactivity concentration of concrete debris, using the approved procedures. By the end of 2012, the authority had confirmed the correctness of our measurement and evaluation results on about 2,600 tons of concrete debris. About 1,800 tons of those cleared concrete were recycled to flat surface subsidence in the NSRI caused by the Great East Japan Earthquake, and also used as base material to construct new buildings and parking lots in the NSRI. (author)

Tight closure and vanishing theorems

International Nuclear Information System (INIS)

Smith, K.E.

2001-01-01

Tight closure has become a thriving branch of commutative algebra since it was first introduced by Mel Hochster and Craig Huneke in 1986. Over the past few years, it has become increasingly clear that tight closure has deep connections with complex algebraic geometry as well, especially with those areas of algebraic geometry where vanishing theorems play a starring role. The purpose of these lectures is to introduce tight closure and to explain some of these connections with algebraic geometry. Tight closure is basically a technique for harnessing the power of the Frobenius map. The use of the Frobenius map to prove theorems about complex algebraic varieties is a familiar technique in algebraic geometry, so it should perhaps come as no surprise that tight closure is applicable to algebraic geometry. On the other hand, it seems that so far we are only seeing the tip of a large and very beautiful iceberg in terms of tight closure's interpretation and applications to algebraic geometry. Interestingly, although tight closure is a 'characteristic p' tool, many of the problems where tight closure has proved useful have also yielded to analytic (L2) techniques. Despite some striking parallels, there had been no specific result directly linking tight closure and L∼ techniques. Recently, however, the equivalence of an ideal central to the theory of tight closure was shown to be equivalent to a certain 'multiplier ideal' first defined using L2 methods. Presumably, deeper connections will continue to emerge. There are two main types of problems for which tight closure has been helpful: in identifying nice structure and in establishing uniform behavior. The original algebraic applications of tight closure include, for example, a quick proof of the Hochster-Roberts theorem on the Cohen-Macaulayness of rings of invariants, and also a refined version of the Brianqon-Skoda theorem on the uniform behaviour of integral closures of powers of ideals. More recent, geometric

The Remote Handled Immobilization Low Activity Waste Disposal Facility Environmental Permits & Approval Plan

Energy Technology Data Exchange (ETDEWEB)

DEFFENBAUGH, M.L.

2000-08-01

The purpose of this document is to revise Document HNF-SD-ENV-EE-003, ''Permitting Plan for the Immobilized Low-Activity Waste Project, which was submitted on September 4, 1997. That plan accounted for the interim storage and disposal of Immobilized-Low Activity Waste at the existing Grout Treatment Facility Vaults (Project W-465) and within a newly constructed facility (Project W-520). Project W-520 was to have contained a combination of concrete vaults and trenches. This document supersedes that plan because of two subsequent items: (1) A disposal authorization that was received on October 25, 1999, in a U. S. Department of Energy-Headquarters, memorandum, ''Disposal Authorization Statement for the Department of Energy Hanford site Low-Level Waste Disposal facilities'' and (2) ''Breakthrough Initiative Immobilized Low-Activity Waste (ILAW) Disposal Alternative,'' August 1999, from Lucas Incorporated, Richland, Washington. The direction within the U. S. Department of Energy-Headquarters memorandum was given as follows: ''The DOE Radioactive Waste Management Order requires that a Disposal authorization statement be obtained prior to construction of new low-level waste disposal facility. Field elements with the existing low-level waste disposal facilities shall obtain a disposal authorization statement in accordance with the schedule in the complex-wide Low-Level Waste Management Program Plan. The disposal authorization statement shall be issued based on a review of the facility's performance assessment and composite analysis or appropriate CERCLA documentation. The disposal authorization shall specify the limits and conditions on construction, design, operations, and closure of the low-level waste facility based on these reviews. A disposal authorization statement is a part of the required radioactive waste management basis for a disposal facility. Failure to obtain a disposal authorization statement

40 CFR 264.228 - Closure and post-closure care.

Science.gov (United States)

2010-07-01

... remaining wastes to a bearing capacity sufficient to support final cover; and (iii) Cover the surface....112 must include both a plan for complying with paragraph (a)(1) of this section and a contingent plan... practicably removed at closure; and (ii) The owner or operator must prepare a contingent post-closure plan...

105-DR Large sodium fire facility soil sampling data evaluation report

International Nuclear Information System (INIS)

Adler, J.G.

1996-01-01

This report evaluates the soil sampling activities, soil sample analysis, and soil sample data associated with the closure activities at the 105-DR Large Sodium Fire Facility. The evaluation compares these activities to the regulatory requirements for meeting clean closure. The report concludes that there is no soil contamination from the waste treatment activities

Construction of Kashiwazaki-Kariwa Nuclear Power Station Results of manufacturing concrete

International Nuclear Information System (INIS)

Morishita, Hideki; Tsuchiya, Yoshimasa; Eguchi, Kiyoshi; Hosaka, Hiroshi

1998-01-01

The construction of Kashiwazaki-Kariwa Nuclear Power Station of Tokyo Electric Power Co., Inc. was completed in July, 1997. Seven nuclear power plants generate about 8.2 million kW, and it is the largest nuclear power station in the world. In the construction, from May, 1980 to August, 1996, the concrete of 2.42 million m 3 for architecture and 1.04 million m 3 for civil engineering, 3.46 million m 3 in total, and the mortar for artificial rock bed of 430,000 m 3 were manufactured and placed. The results of manufacturing concrete from beginning to finish are shown. The specification of concrete was different for No. 1 plant, No. 2 and 5 plants, No. 3 and 4 plants and No. 6 and 7 plants. As to the mixing of concrete, the specification and the materials used are reported. The features of the facilities for manufacturing concrete are explained. The flowchart of the quality control of materials and concrete is shown. The material testing of cement and aggregate, the test of water quality and the material testing of admixtures were carried out. As for concrete, the weight of unit volume, slump, air quantity, concrete temperature, chloride content, strength and alkali reactivity were examined. (K.I.)

TRANSPARENT CONCRETE

OpenAIRE

Sandeep Sharma*, Dr. O.P. Reddy

2017-01-01

Transparent concrete is the new type of concrete introduced in todays world which carries special property of light transmitting due to presence of light Optical fibres. Which is also known as translucent concrete or light transmitting concrete, it is achieved by replacing coarse aggregates with transparent alternate materials (Optical fibres). The binding material in transparent concrete may be able to transmit light by using clear resins the concrete mix. The concrete used in industry in pr...

Post-operative analgesic requirement in non-closure and closure of peritoneum during open appendectomy

International Nuclear Information System (INIS)

Khan, A.W.; Maqsood, R.; Saleem, M.M.

2017-01-01

To compare the mean post-operative analgesic requirement in non-closure and closure of peritoneum during open appendectomy. Study Design: Randomized controlled trial. Place and Duration of Study: Department of General Surgery Combined Military Hospital Quetta, from 1st August 2014 to 30th April 2015. Material and Methods: A total of 60 patients were included in this study and were divided into two groups of 30 each. Patients in group A underwent open appendectomy with closure of peritoneum while patients in group B had non-closure of peritoneum during the same procedure. Post-operatively, pain severity was assessed on visual analogue scale (VAS) numeric pain distress scale. On presence of VAS numeric pain distress scale between 5 to 7, intramuscular (IM) diclofenac sodium was given and on score >7, intravascular (IV) tramadol was given. The final outcome was measured at day 0 and day 1. Results: Pain score and analgesic requirements were significantly less in non-closure group than closure group on day 0 and day 1, showing statistically significant difference between the two groups. Conclusion: Mean post-operative analgesic requirement is significantly less in non-closure group as compared to closure group during open appendectomy. (author)

Annual Report RCRA Post-Closure Monitoring and Inspections for Corrective Action Unit 91: Area 3 U-3fi Injection Well, Nevada Test Site, Nevada, for the Period October 2001 - October 2002

International Nuclear Information System (INIS)

Richardson, G.

2003-01-01

This annual monitoring and inspection report provides an analysis and summary for site inspections, meteorological information, and neutron soil moisture monitoring data obtained at the U-3fi Injection Well during the October 2001 to October 2002 period. The U-3fi Injection Well is located in Area 3 of the Nevada Test Site (NTS), Nye County, Nevada. Inspections of the Area 3 U-3fi Injection Well are conducted to determine and document the physical condition of the concrete pad, facilities, and any unusual conditions that could impact the proper operation of the waste disposal unit closure. The objective of the neutron logging is to monitor the soil moisture conditions along the 128-meter (m) (420-feet [ft]) ER3-3 monitoring well and detect changes that may be indicative of moisture movement in the regulated interval extending between 73 to 82 m (240 to 270 ft)

Primary closure after carotid endarterectomy is not inferior to other closure techniques.

Science.gov (United States)

Avgerinos, Efthymios D; Chaer, Rabih A; Naddaf, Abdallah; El-Shazly, Omar M; Marone, Luke; Makaroun, Michel S

2016-09-01

Primary closure after carotid endarterectomy (CEA) has been much maligned as an inferior technique with worse outcomes than in patch closure. Our purpose was to compare perioperative and long-term results of different CEA closure techniques in a large institutional experience. A consecutive cohort of CEAs between January 1, 2000, and December 31, 2010, was retrospectively analyzed. Closure technique was used to divide patients into three groups: primary longitudinal arteriotomy closure (PRC), patch closure (PAC), and eversion closure (EVC). End points were perioperative events, long-term strokes, and restenosis ≥70%. Multivariate regression models were used to assess the effect of baseline predictors. There were 1737 CEA cases (bilateral, 143; mean age, 71.4 ± 9.3 years; 56.2% men; 35.3% symptomatic) performed during the study period with a mean clinical follow-up of 49.8 ± 36.4 months (range, 0-155 months). More men had primary closure, but other demographic and baseline symptoms were similar between groups. Half the patients had PAC, with the rest evenly distributed between PRC and EVC. The rate of nerve injury was 2.7%, the rate of reintervention for hematoma was 1.5%, and the length of hospital stay was 2.4 ± 3.0 days, with no significant differences among groups. The combined stroke and death rate was 2.5% overall and 3.9% and 1.7% in the symptomatic and asymptomatic cohort, respectively. Stroke and death rates were similar between groups: PRC, 11 (2.7%); PAC, 19 (2.2%); EVC, 13 (2.9%). Multivariate analysis showed baseline symptomatic disease (odds ratio, 2.4; P = .007) and heart failure (odds ratio, 3.1; P = .003) as predictors of perioperative stroke and death, but not the type of closure. Cox regression analysis demonstrated, among other risk factors, no statin use (hazard ratio, 2.1; P = .008) as a predictor of ipsilateral stroke and severe (glomerular filtration rate <30 mL/min/1.73 m(2)) renal insufficiency (hazard ratio, 2.6; P�

Numerical homogenization of concrete microstructures without explicit meshes

International Nuclear Information System (INIS)

Sanahuja, Julien; Toulemonde, Charles

2011-01-01

Life management of electric hydro or nuclear power plants requires to estimate long-term concrete properties on facilities, for obvious safety and serviceability reasons. Decades-old structures are foreseen to be operational for several more decades. As a large number of different concrete formulations are found in EDF facilities, empirical models based on many experiments cannot be an option for a large fleet of power plant buildings. To build predictive models, homogenization techniques offer an appealing alternative. To properly upscale creep, especially at long term, a rather precise description of the microstructure is required. However, the complexity of the morphology of concrete poses several challenges. In particular, concrete is formulated to maximize the packing density of the granular skeleton, leading to aggregates spanning several length scales with small inter particle spacings. Thus, explicit meshing of realistic concrete microstructures is either out of reach of current meshing algorithms or would produce a number of degrees of freedom far higher than the current generic FEM codes capabilities. This paper proposes a method to deal with complex matrix-inclusions microstructures such as the ones encountered at the mortar or concrete scales, without explicitly meshing them. The microstructure is superimposed to an independent mesh, which is a regular Cartesian grid. This inevitably yields so called 'gray elements', spanning across multiple phases. As the reliability of the estimate of the effective properties highly depends on the behavior affected to these gray elements, special attention is paid to them. As far as the question of the solvers is concerned, generic FEM codes are found to lack efficiency: they cannot reach high enough levels of discretization with classical free meshes, and they do not take advantage of the regular structure of the mesh. Thus, a specific finite differences/finite volumes solver has been developed. At first, generic off

Optimization of concrete composition in radioactive waste management

International Nuclear Information System (INIS)

Plecas, I.; Peric, A.

1995-01-01

Low and intermediate level waste represents 95% of the total wastes that is conditioned into special concrete containers. Since these containers are to protect radioactive waste safely for about 300 years, the selection and precise control of physical and mechanical characteristics of materials is very important. After volume reduction and valuable components recovery, waste materials have to be conditioned for transport, storage and disposal. Conditioning is the waste management step in which radioactive wastes are immobilized and packed. The immobilization processes involve conversation of the wastes to solid forms that reduce the potential for migration or dispersion of radionuclides from the wastes by natural processes during storage, transport and disposal. The immobilization processes involve the use of various matrices of nonradioactive materials, such as concrete, to fix the wastes as monoliths, usually directly in the waste containers used for subsequent handling. In this paper an optimization of concrete container composition, used for storing radioactive waste from nuclear power plants, is presented. Optimization was performed on the composition of the concrete that is used in the container production. In experiments, the authors tried to obtain the best mechanical characteristics of the concrete, varying the weight percentage of the granulate due to its diameter, water-to-cement ratios and type of the cements that were used in preparing the concrete container formulation. Concrete containers, that were optimized in the manner described in this paper, will be in used for the radioactive waste materials final disposal, using the concept of the engineer trench system facilities

Public perception of odour and environmental pollution attributed to MSW treatment and disposal facilities: A case study

International Nuclear Information System (INIS)

De Feo, Giovanni; De Gisi, Sabino; Williams, Ian D.

2013-01-01

Highlights: ► Effects of closing MSW facilities on perception of odour and pollution studied. ► Residents’ perception of odour nuisance considerably diminished post closure. ► Odour perception showed an association with distance from MSW facilities. ► Media coverage increased knowledge about MSW facilities and how they operate. ► Economic compensation possibly affected residents’ views and concerns. - Abstract: If residents’ perceptions, concerns and attitudes towards waste management facilities are either not well understood or underestimated, people can produce strong opposition that may include protest demonstrations and violent conflicts such as those experienced in the Campania Region of Italy. The aim of this study was to verify the effects of the closure of solid waste treatment and disposal facilities (two landfills and one RDF production plant) on public perception of odour and environmental pollution. The study took place in four villages in Southern Italy. Identical questionnaires were administered to residents during 2003 and after the closure of the facilities occurred in 2008. The residents’ perception of odour nuisance considerably diminished between 2003 and 2009 for the nearest villages, with odour perception showing an association with distance from the facilities. Post closure, residents had difficulty in identifying the type of smell due to the decrease in odour level. During both surveys, older residents reported most concern about the potentially adverse health impacts of long-term exposure to odours from MSW facilities. However, although awareness of MSW facilities and concern about potentially adverse health impacts varied according to the characteristics of residents in 2003, substantial media coverage produced an equalisation effect and increased knowledge about the type of facilities and how they operated. It is possible that residents of the village nearest to the facilities reported lower awareness of and concern about

Public perception of odour and environmental pollution attributed to MSW treatment and disposal facilities: A case study

Energy Technology Data Exchange (ETDEWEB)

De Feo, Giovanni, E-mail: g.defeo@unisa.it [Department of Industrial Engineering, University of Salerno, via Ponte don Melillo 1, 84084 Fisciano (Italy); De Gisi, Sabino [Department of Industrial Engineering, University of Salerno, via Ponte don Melillo 1, 84084 Fisciano (Italy); Williams, Ian D. [Waste Management Research Group, Faculty of Engineering and the Environment, University of Southampton, Highfield, Southampton SO17 1BJ (United Kingdom)

2013-04-15

Highlights: ► Effects of closing MSW facilities on perception of odour and pollution studied. ► Residents’ perception of odour nuisance considerably diminished post closure. ► Odour perception showed an association with distance from MSW facilities. ► Media coverage increased knowledge about MSW facilities and how they operate. ► Economic compensation possibly affected residents’ views and concerns. - Abstract: If residents’ perceptions, concerns and attitudes towards waste management facilities are either not well understood or underestimated, people can produce strong opposition that may include protest demonstrations and violent conflicts such as those experienced in the Campania Region of Italy. The aim of this study was to verify the effects of the closure of solid waste treatment and disposal facilities (two landfills and one RDF production plant) on public perception of odour and environmental pollution. The study took place in four villages in Southern Italy. Identical questionnaires were administered to residents during 2003 and after the closure of the facilities occurred in 2008. The residents’ perception of odour nuisance considerably diminished between 2003 and 2009 for the nearest villages, with odour perception showing an association with distance from the facilities. Post closure, residents had difficulty in identifying the type of smell due to the decrease in odour level. During both surveys, older residents reported most concern about the potentially adverse health impacts of long-term exposure to odours from MSW facilities. However, although awareness of MSW facilities and concern about potentially adverse health impacts varied according to the characteristics of residents in 2003, substantial media coverage produced an equalisation effect and increased knowledge about the type of facilities and how they operated. It is possible that residents of the village nearest to the facilities reported lower awareness of and concern about

Using natural local materials for developing special radiation shielding concretes, and deduction of its shielding characteristics

International Nuclear Information System (INIS)

Kharita, M. H.; Takeyeddin, M.; Al-Nassar, M.; Yousef, S.

2006-06-01

Concrete is considered as the most important material to be used for radiation shielding in facilities contain radioactive sources and radiation generating machines. The concrete shielding properties may vary depending on the construction of the concrete, which is highly relative to the composing aggregates i.e. aggregates consist about 70 - 80% of the total weight of normal concrete. In this project tow types of concrete used in Syria (in Damascus and Aleppo) had been studied and their shielding properties were defined for gamma ray from Cs-137 and Co-60 sources, and for neutrons from Am-Be source. About 10% reduction in HVL was found in the comparison between the tow concrete types for both neutrons and gammas. Some other types of concrete were studied using aggregates from different regions in Syria, to improve the shielding properties of concrete, and another 10% of reduction was achieved in comparison with Damascene concrete (20% in comparison with the concrete from Aleppo) for both neutrons and gamma rays. (author)

Corium Spreading Over Concrete: The Vulcano VE-U7 and VE-U8 Tests

International Nuclear Information System (INIS)

Journeau, Christophe; Boccaccio, Eric; Fouquart, Pascal; Jegou, Claude; Piluso, Pascal

2002-01-01

Two experiments have been performed in the VULCANO facility in which prototypic corium has been spread over concrete. In the VE-U7 test, a mixture representative of what can be expected at the opening of EPR reactor-pit gate has been spread on siliceous concrete and on a reference channel in inert refractory ceramic. The spreading progression was not much affected by the presence of concrete and sparging gases. In the VE-U8 test, a UO 2 -ZrO 2 mixture, prototypic of in-vessel corium, has been spread over a lime-siliceous concrete. Although residual power was not simulated in this experiment, up to 2 cm of concrete have been eroded during the test. Results in terms of spreading behaviour, effects of gases, concrete erosion and thermal attack are presented and discussed. (authors)

SKI and SSI's recommendations to the government concerning long-term responsibility after closure of a repository for spent nuclear fuel

International Nuclear Information System (INIS)

Paeivioe Jonsson, Josefin

2008-01-01

Many activities will cease at the closure of a repository, but not responsibilities. The candidate municipalities in Sweden expressed concern about who will take over after the implementer is released from responsibility for the facility. The government thus commissioned SKI (Swedish Nuclear Power Inspectorate) and SSI (Swedish Radiation Protection Authority) to review the legal obligations of institutional players as laid out today in legislation in Sweden. After closure of the repository in about 100 years there will be post-closure monitoring, possibly for a few hundred years. This will be a part of the conditions on SKB (Swedish Nuclear Fuel and Waste Management Company) which will be set out at the time. Some activities will end at the closure of the facility but monitoring and safeguards obligations may continue. The exact nature of this monitoring and safeguard work needs to be discussed and agreed upon. With the proposed approach most of the liabilities rest with the state in the long term, the waste producers only have liabilities in the short term but their decisions could have big impacts on long term liabilities

International symposium on seismic evaluation of existing nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

Orbovic, N.; Bouchon, M. [Institut de Radioprotection et de Surete Nucleaire, IRSN, 92 - Fontenay aux Roses (France); Vendel, J.; Gelain, T. [IRSN/DPEA/SERAC, 91 - Gif sur Yvette (France)

2003-10-01

Below are summarized the works of Mrs N. Orbovic and M. Bouchon; B. Stojadinovic, N. Orbovic, M.Bouchon and J.T. Wiley; M. Bouchon, N. Orbovic and B. Foure; T. Gelain, F. Gensdarmes, R. Sestier-Carlin, J. Vendel and M. Bouchon dealing respectively with: 1) seismic assessment of existing nuclear facility: a case study position of the IRSN 2) static and dynamic evaluation of an existing nuclear facility reinforced concrete frame structure 3) experimental study of cracking of low-rise reinforced concrete shear walls 4) aeraulics study of crack networks on low-rise reinforced concrete walls subject to static cycling loading. The summaries are followed: 1) The Institute for Radioprotection and Nuclear Safety (IRSN) has the task of evaluating the seismic safety of existing building inventory in French nuclear facilities. Some of the existing structures are reinforced concrete frame buildings with masonry or reinforced concrete in-fill walls built during the 1960's, 70's and 80's following different building codes and seismic input data applicable at the time of construction. The studied building is a laboratory that was built in 1962. The building is composed of three different and independent blocks. The structure is a reinforced concrete frame with masonry in-fills and few stiff concrete elements. Moreover, the building contains an independent massive concrete cell. Structurally, the building has a number of irregularities. From the detailing standpoint, well known deficiencies of low-ductile reinforced concrete structures are evident especially, a lack of transverse reinforcement in the structural joints of the frame. A number of sensitivity studies were conducted using 2D and 3D linear models to evaluate seismic demand. Ambient vibration and regional earthquake records were used to check the soil nature and the existence or not of a possible site effect around the installation as well as to characterize the dynamic behavior of the building. 3D models

Integrating scientific results for a post-closure safety demonstration

International Nuclear Information System (INIS)

Taylor, E.C.; Ramspott, L.D.; Sinnock, S.; Sprecher, W.M.

1994-01-01

The U.S. Department of Energy (DOE) is developing a nuclear waste management system that will accept high-level radioactive waste, transport it, store it, and ultimately emplace it in a deep geologic repository. The key activity now is determining whether Yucca Mountain, Nevada is suitable as a site for the repository. If so, the crucial technological advance will be the demonstration that disposal of nuclear waste will be safe for thousands of years after closure. Recent regulatory, legal, and scientific developments imply that the safety demonstration must be simple. The scientific developments taken together support a simple set of hypotheses that constitute a post-closure safety argument for a repository at Yucca Mountain. If the understanding of Yucca Mountain hydrology presented in the Site Characterization Plan proves correct, then these hypotheses might be confirmed by combining results of Surface-Based Testing with early testing results in the Exploratory Studies Facility

Interim Safety Basis for Fuel Supply Shutdown Facility

International Nuclear Information System (INIS)

BENECKE, M.W.

2000-01-01

This ISB, in conjunction with the IOSR, provides the required basis for interim operation or restrictions on interim operations and administrative controls for the facility until a SAR is prepared in accordance with the new requirements or the facility is shut down. It is concluded that the risks associated with tha current and anticipated mode of the facility, uranium disposition, clean up, and transition activities required for permanent closure, are within risk guidelines

How Concrete is Concrete

Directory of Open Access Journals (Sweden)

Koeno Gravemeijer

2010-07-01

Full Text Available If we want to make something concrete in mathematics education, we are inclined introduce, what we call, ‘manipulatives’, in the form of tactile objects or visual representations. If we want to make something concrete in a everyday-life conversation, we look for an example. In the former, we try to make a concrete model of our own, abstract, knowledge; in the latter, we try to find an example that the others will be familiar with. This article first looks at the tension between these two different ways of making things concrete. Next another role of manipulatives, will be discussed, namely that of means for scaffolding and communication. In this role, manipulatives may function as means of support in a process that aims at helping students to build on their own thinking while constructing more sophisticated mathematics

How Concrete is Concrete

OpenAIRE

Koeno Gravemeijer

2010-01-01

If we want to make something concrete in mathematics education, we are inclined introduce, what we call, ‘manipulatives’, in the form of tactile objects or visual representations. If we want to make something concrete in a everyday-life conversation, we look for an example. In the former, we try to make a concrete model of our own, abstract, knowledge; in the latter, we try to find an example that the others will be familiar with. This article first looks at the tension between these two diff...

Thermal and hydraulic properties of the concrete used in the ILW repository of El Cabril (Spain). Preliminary laboratory tests

International Nuclear Information System (INIS)

Villar, Maria Victoria

2012-01-01

This work is a contribution to the understanding of the behaviour of concrete barriers in surface (low and intermediate-level) waste disposal facilities, in particular in the Spanish disposal facility of El Cabril, where the waste containers are placed inside concrete cells. The durability of concrete and its mechanical properties are intrinsically bound to moisture transport effects, especially when it is subjected to repeated wetting and drying regimes, and that is why a detailed thermo-hydraulic characterisation is necessary to model its behaviour. The concrete used in this experimental work has a characteristic strength of 350 kp/cm 2 and uses ordinary Portland cement, resistant to sulphates and seawater, with a water/cement ratio of 0.43. Its average pore size is 0.03 micrometer. In addition to the determination of the thermal conductivity of concrete as a function of water content, a hydraulic characterisation - including determination of saturated permeability, permeability to gas for different degrees of saturation and water retention curves - has been performed

Washeteria closures, infectious disease and community health in rural Alaska: a review of clinical data in Kivalina, Alaska

Directory of Open Access Journals (Sweden)

Timothy K. Thomas

2013-08-01

Full Text Available Background. Kivalina is a northwest Alaska barrier island village of 400 people vulnerable to storm surges exacerbated recently by delayed winter sea and shore ice formation. The village has no in-home piped water or sewage; the “washeteria” is the only structure providing public showers, laundry facilities and flush toilets. In October 2004, a storm damaged the washeteria septic system resulting in prolonged facility closures. We assessed rates of gastrointestinal, respiratory and skin infections potentially impacted by prolonged washeteria closures. Methods. We obtained washeteria closure dates from 2003 to July 2009 and defined >7 day closure as prolonged. We received de-identified data on all Kivalina clinic visits from 2003 to 2009 and selected visits with ICD-9 diagnosis codes for respiratory, skin, or gastrointestinal infection; subsequent same patient/same illness-category visits within 14 days were excluded. We compared annual visit rates, for all ages combined, before (2003–2004 and after (2005–2009 the “2004” storm. Results. The washeteria had prolonged closures for 34 days (4.7% in the 2 years 2003–2004 and 864 days (51.7% between January 2005 and July 2009. Closures ranged from 8 to 248 days. Respiratory infection rates declined significantly from 1.32 visits/person/year in the 2003–2004 period to 0.99 visits/person/year in the 2005–2009 period. There was a significant increase in skin infection rates after 2004, peaking at 0.28 visits/person/year in 2007 and then declining significantly to 0.15 visits/person/year in 2009. Gastrointestinal infection rates remained stable and low throughout (average: 0.05 visits/person/year. No temporal association was observed between respiratory, gastrointestinal or skin infection rates and prolonged washeteria closures. Conclusion. The Kivalina washeteria was closed frequently and for extended periods between 2005 and 2009. Initial closures possibly resulted in increased skin