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Sample records for emplaced spent unreprocessed

Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository: accident event analysis and mechanical failure probabilities. Final report

International Nuclear Information System (INIS)

Bhaskaran, G.; McCleery, J.E.

1979-10-01

This report provides support in developing an accident prediction event tree diagram, with an analysis of the baseline design concept for the retrieval of emplaced spent unreprocessed fuel (SURF) contained in a degraded Canister. The report contains an evaluation check list, accident logic diagrams, accident event tables, fault trees/event trees and discussions of failure probabilities for the following subsystems as potential contributors to a failure: (a) Canister extraction, including the core and ram units; (b) Canister transfer at the hoist area; and (c) Canister hoisting. This report is the second volume of a series. It continues and expands upon the report Retrieval System for Emplaced Spent Unreprocessed Fuel (SURF) in Salt Bed Depository: Baseline Concept Criteria Specifications and Mechanical Failure Probabilities. This report draws upon the baseline conceptual specifications contained in the first report
Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository. Baseline concept criteria specifications and mechanical failure probabilities

International Nuclear Information System (INIS)

Hudson, E.E.; McCleery, J.E.

1979-05-01

One of the integral elements of the Nuclear Waste Management Program is the material handling task of retrieving Canisters containing spent unreprocessed fuel from their emplacement in a deep geologic salt bed Depository. A study of the retrieval concept data base predicated this report. In this report, alternative concepts for the tasks are illustrated and critiqued, a baseline concept in scenario form is derived and basic retrieval subsystem specifications are presented with cyclic failure probabilities predicted. The report is based on the following assumptions: (a) during retrieval, a temporary radiation seal is placed over each Canister emplacement; (b) a sleeve, surrounding the Canister, was initially installed during the original emplacement; (c) the emplacement room's physical and environmental conditions established in this report are maintained while the task is performed
Risk analysis methodology for unreprocessed spent fuel disposal in bedded salt

International Nuclear Information System (INIS)

Pepping, R.E.; Chu, M.S.Y.; Cranwell, R.M.

1982-01-01

In accordance with the decision to defer the reprocessing of commercially generated spent fuel, we are investigating the implications on risk of direct disposal of spent fuel assemblies. To the extent possible, we are using the methodology developed at Sandia for the NRC to evaluate risks from the disposal of wastes from reprocessing of spent fuel. This allows direct comparison of the risks calculated for the two waste forms. A number of differences between the two waste forms with implications on risk have been identified and investigation of their effects has begun. Among these are the presence of gases and additional plutonium and uranium isotopes, the potential for differing leach behavior, and the difference in the decay heat source which determines the overall thermomechanical response of the host media. We have analyzed a number of scenarios for a hypothetical geologic repository that have been identified as important contributors to risk from the disposal of both reprocessed and unreprocessed spent fuel. For each scenario, we employ the Groundwater Transport, Pathways to Man, and Dosimetry and Health Effects models of the High Level Waste Methodology. Risks are compared for the reprocessed and unreprocessed spent fuel wastes and the effects of uncertainty in the parameters of the various models are compared
Potential sites for a spent unreprocessed fuel facility (SURFF), southwesten part of the Nevada Test Site

International Nuclear Information System (INIS)

Hoover, D.L.; Eckel, E.B.; Ohl, J.P.

1978-01-01

In the absence of specific criteria, the topography, geomorphology, and geology of Jackass Flats and vicinity in the southwestern part of the Nevada Test Site are evaluated by arbitrary guidelines for a Spent Unreprocessed Fuel Facility. The guidelines include requirements for surface slopes of less than 5%, 61 m of alluvium beneath the site, an area free of active erosion or deposition, lack of faults, a minimum area of 5 km 2 , no potential for flooding, and as many logistical support facilities as possible. The geology of the Jackass Flats area is similar to the rest of the Nevada Test Site in topographic relief (305-1,200 m), stratigraphy (complexly folded and faulted Paleozoic sediments overlain by Tertiary ash-flow tuffs and lavas overlain in turn by younger alluvium), and structure (Paleozoic thrust faults and folds, strike-slip faults, proximity to volcanic centers, and Basin and Range normal faults). Of the stratigraphic units at the potential Spent Unreprocessed Fuel Facility site in Jackass Flats, only the thickness and stability of the alluvium are of immediate importance. Basin and Range faults and a possible extension of the Mine Mountain fault need further investigation. The combination of a slope map and a simplified geologic and physiographic map into one map shows several potential sites for a Spent Unreprocessed Fuel Facility in Jackass Flats. The potential areas have slopes of less than 5% and contain only desert pavement or segmented pavement--the two physiographic categories having the greatest geomorphic and hydraulic stability. Before further work can be done, specific criteria for a Spent Unreprocessed Fuel Facility site must be defined. Following criteria definition, potential sites will require detailed topographic and geologic studies, subsurface investigations (including geophysical methods, trenching, and perhaps shallow drilling for faults in alluvium), detailed surface hydrologic studies, and possibly subsurface hydrologic studies
Spent unreprocessed fuel (SURF) facility evaluation plan of the alternative storage concepts

International Nuclear Information System (INIS)

Berry, S.M.

1978-01-01

Concepts were evaluated for the storage of unreprocessed spent fuel in a retrievable surface storage facility. This document provides a systematic format for making a concept selection from the seven alternative concepts presented in RHO-LD-2. Results of the evaluation was that the Drywell concept was rated highest with the Water Basin Concept and the Sealed Storage Cask concept with multiple canisters of SURF coming in a close second and third
Handling final storage of unreprocessed spent nuclear fuel

International Nuclear Information System (INIS)

1978-01-01

The present second report from KBS describes how the safe final storage of spent unreprocessed nuclear fuel can be implemented. According to the Swedish Stipulation Law, the owner must specify in which form the waste is to be stored, how final storage is to be effected, how the waste is to be transported and all other aspects of fuel handling and storage which must be taken into consideration in judging whether the proposed final storage method can be considered to be absolutely safe and feasible. Thus, the description must go beyond general plans and sketches. The description is therefore relatively detailed, even concerning those parts which are less essential for evaluating the safety of the waste storage method. For those parts of the handling chain which are the same for both alternatives of the Stipulation Law, the reader is referred in some cases to the first report. Both of the alternatives of the Stipulation Law may be used in the future. Handling equipment and facilities for the two storage methods are so designed that a combination in the desired proportions is practically feasible. In this first part of the report are presented: premises and data, a description of the various steps of the handling procedure, a summary of dispersal processes and a safety analysis. (author)
National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Volume I. Conceptual design report

International Nuclear Information System (INIS)

1978-12-01

In February 1976, the Energy Research and Development Administration (ERDA), now the Department of Energy (DOE), established a National Waste Terminal Storage (NWTS) program. As a part of this program, two parallel conceptual design efforts were initiated in January 1977. One was for deep geologic storage, in domed salt, of high level waste resulting from the reprocessing of spent fuel. The other was for deep geologic storage of unreprocessed spent fuel in bedded salt. These two concepts are identified as NWTS Repository 1 and Repository 2, respectively. Repository 2 (NWTSR2) is the concept which is covered by this Conceptual Design Report. Volume I of the conceptual design report contains the following information: physical description of the report; project purpose and justification; principal safety, fire, and health hazards; environmental impact considerations; quality assurance considerations; assessment of operational interfaces; assessment of research and development interfaces; project schedule; proposed method of accomplishment; summary cost estimate; and outline specifications. The conceptual design for Repository 2 was developed in sufficient detail to permit determination of scope, engineering feasibility, schedule, and cost estimates, all of which are necessary for planning and budgeting the project
Risk analysis methodology for spent fuel repositories in bedded salt: methodlogy summary and differences between spent fuel and high level wastes

International Nuclear Information System (INIS)

Pepping, R.E.; Chu, M.S.

1981-06-01

In the absence of spent fuel reprocessing plans, unreprocessed spent fuel has become a candidate waste form for geologic disposal. In order to understand the public health risks from such disposal and to gain insights into the factors that influence them, a methodology is needed to combine the effects of site geology and hydrology, physical and chemical properties of the waste form, and the details of the engineering design. This report outlines such a methodology which the authors currently are applying to the analysis of unreprocessed spent fuel disposal. The methodology is the same methodology as was developed to describe the risks from geologic disposal of wastes from reprocessed spent fuel. The difference between spent fuel wastes and wastes from reprocessing that may affect the applicability of the methodology are highlighted
National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Quality assurance program for licensing

International Nuclear Information System (INIS)

1978-12-01

A National Waste Terminal Storage Repository, in bedded salt, for spent unreprocessed fuel is the subject of a conceptual design project which began in January 1977. This volume presents a preliminary quality assurance program to guide the license applicant in developing a detailed program that will be compatible with anticipated National Waste Terminal Storage (NWTSR2) contracting arrangements and provide the documentation required by regulatory bodies. This QA program is designed to provide confidence that the quality-related activities pertaining to safety-related structures, systems, and components will be identified and controlled. Specific responsibilities for quality-related activities are documented and assigned to personnel and organizations for the major phases of facility design and construction. These responsibilities encompass a broad range of activities and are addressed in this preliminary program. The quality assurance program elements are organized and discussed herein as follows: (1) quality assurance during design and construction; (2) the applicant (DOE); (3) siting contractor; (4) architect/engineer; (5) project field management; and (6) operations contractor
Thermoelastic analysis of spent fuel and high level radioactive waste repositories in salt. A semi-analytical solution

International Nuclear Information System (INIS)

St John, C.M.

1977-04-01

An underground repository containing heat generating, High Level Waste or Spent Unreprocessed Fuel may be approximated as a finite number of heat sources distributed across the plane of the repository. The resulting temperature, displacement and stress changes may be calculated using analytical solutions, providing linear thermoelasticity is assumed. This report documents a computer program based on this approach and gives results that form the basis for a comparison between the effects of disposing of High Level Waste and Spent Unreprocessed Fuel
Emplacement technology for the direct disposal of spent fuel into deep vertical boreholes

International Nuclear Information System (INIS)

Bollingerfehr, W.; Filbert, W.; Wehrmann, J.

2008-01-01

In the early sixties it was decided to investigate salt formations on its suitability to host heat generating radioactive waste in Germany. In the reference repository concept consequently the emplacement of vitrified waste canisters in deep vertical boreholes inside a salt mine was considered whereas spent fuel should be disposed of in self shielding casks (type POLLUX) in horizontal drifts. The POLLUX casks, 65 t heavy carbon steel casks, will be laid down on the floor of a horizontal drift in one of the disposal zones to be constructed in the salt dome at the 870 m level. The space between casks and drift walls will be backfilled with crushed salt. The transport, the handling und the emplacement of POLLUX casks were subject of successfully performed demonstration and in situ tests in the nineties and resulted in an adjustment of the atomic law. The borehole disposal concept comprises the emplacement of unshielded canisters with vitrified HLW in boreholes with a diameter of 60 cm and a depth of up to 300 m. In order to facilitate the fast encapsulation of the waste canister by the host rock (rock salt), no lining of the boreholes is planned. With regard to harmonize and optimize the emplacement technology for both categories of packages (vitrified waste and spent fuel) alternatives were developed. In this context the borehole emplacement technique for consolidated spent fuel as already foreseen for high-level reprocessing waste was reconsidered. This review resulted in the design of a new disposal package, a fuel rod canister (type 'BSK 3'), and an appropriate modified transport and emplacement technology. This concept (called BSK 3-concept) provides the following optimization possibilities: (i) A new steel canister of the same diameter (43 cm) as the standardized HLW canisters applied for high-level waste and compacted technological waste from reprocessing abroad can be filled with fuel rods of 3 PWR or 9 BWR fuel assemblies. (II) The standardized canister
National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Special study No. 1. 25-year retrievability

International Nuclear Information System (INIS)

1978-12-01

National Waste Terminal Storage Repository 2 (NWTSR2), in bedded salt, for spent unreprocessed fuel is the subject of a conceptual design project which began in January 1977. In the base conceptual design the spent fuel is to be stored in a retrievable mode for 5 years. This report presents the results of a separate study on 25-year retrievability with open storage rooms. Drawings prepared for 25-year retrievability are only those which differ from the companion 5-year retrievability drawings which comprise Volume II of the Conceptual Design Report (CDR). NWTSR2 capital facilities originally were to be constructed in phases. Phase I was to include those facilities required for the first 5 years of operation, which were hypothesized to be from 1985 through 1990; Phase II was to include facilities required from 1991 until the capacity of the 2000-acre mine was reached. These dates are used as the basis for conceptual design and for Special Study No. 1 but do not represent actual schedule dates for the repository. The terms Phase I and Phase II are used throughout the drawings and test, so they need to be understood. However, it is not now intended to phase the capital construction because of the relatively small effort represented by Phase II surface construction. An optional future expansion of the mine, Phase III, would add 1000 acres on the end of the mine opposite the shaft pillar. The optional Phase III expansion would commence operation upon completion of Phase II operations. The drawings and design comparisons for Phase III mine operations appear only in this report
Monte-Carlo based comparison of the personal dose for emplacement scenarios of spent nuclear fuel casks in generic deep geological repositories

Energy Technology Data Exchange (ETDEWEB)

Suarez, Hector Sauri; Becker, Franz; Metz, Volker [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Nuclear Waste Disposal (INE); Pang, Bo [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Nuclear Waste Disposal (INE); Shenzhen Univ. (China). College of Physics and Energy

2017-06-15

In the operational phase of a deep geological disposal facility for high-level nuclear waste, the radiation field in the vicinity of a waste cask is influenced by the backscattered radiation of the surrounding walls of the emplacement drift. For a comparison of disposal of spent nuclear fuel in various host rocks, it is of interest to investigate the influence of the surrounding materials on the radiation field and the personal radiation exposure. In this generic study individual dosimetry of personnel involved in emplacement of casks with spent nuclear fuel in drifts in rock salt and in a clay formation was modelled.
Pacific Northwest Laboratory (PNL) spent fuel transportation and handling facility models

Energy Technology Data Exchange (ETDEWEB)

Andrews, W.B.; Bower, J.C.; Burnett, R.A.; Engel, R.L.; Rolland, C.W.

1979-09-01

A spent fuel logistics study was conducted in support of the US DOE program to develop facilities for preparing spent unreprocessed fuel from commercial LWRs for geological storage. Two computerized logistics models were developed. The first one was the site evaluation model. Two studies of spent fuel handling facility and spent fuel disposal facility siting were completed; the first postulates a single spent fuel handling facility located at any of six DOE laboratory sites, while the second study examined siting strategies with the spent fuel repository relative to the spent fuel handling facility. A second model to conduct storage/handling facility simulations was developed. (DLC)
Pacific Northwest Laboratory (PNL) spent fuel transportation and handling facility models

International Nuclear Information System (INIS)

Andrews, W.B.; Bower, J.C.; Burnett, R.A.; Engel, R.L.; Rolland, C.W.

1979-09-01

A spent fuel logistics study was conducted in support of the US DOE program to develop facilities for preparing spent unreprocessed fuel from commercial LWRs for geological storage. Two computerized logistics models were developed. The first one was the site evaluation model. Two studies of spent fuel handling facility and spent fuel disposal facility siting were completed; the first postulates a single spent fuel handling facility located at any of six DOE laboratory sites, while the second study examined siting strategies with the spent fuel repository relative to the spent fuel handling facility. A second model to conduct storage/handling facility simulations was developed
Handling of final storage of unreprocessed spent nuclear fuel

International Nuclear Information System (INIS)

1978-01-01

In this report the various facilities incorporated in the proposed handling chain for spent fuel from the power stations to the final repository are discribed. Thus the geological conditions which are essential for a final repository is discussed as well as the buffer and canister materials and how they contribute towards a long-term isolation of the spent fuel. Furthermore one chapter deals with leaching of the deposited fuel in the event that the canister is penetrated as well as the transport mechanisms which determine the migration of the radioactive substances through the buffer material. The dispersal processes in the geosphere and the biosphere are also described together with the transfer mechanisms to the ecological systems as well as radiation doses. Finally a summary is given of the safety analysis of the proposed method for the handling and final storage of the spent fuel. (E.R.)
German Approach to Spent Fuel Management

International Nuclear Information System (INIS)

Jussofie, A.; Graf, R.; Filbert, W.

2010-01-01

The management of spent fuel was based on two powerful columns until 30 June 2005, i. e. reprocessing and direct disposal. After this date any delivery of spent fuel to reprocessing plants was prohibited so that the direct disposal of unreprocessed spent fuel is the only available option in Germany today. The main steps of the current concept are: (i) Intermediate storage of spent fuel, which is the only step in practice. After the first cooling period in spent fuel storage pools it continues into cask-receiving dry storage facilities. Identification of casks, 'freezing' of inventories in terms of continuity of knowledge, monitoring the access to spent fuel, verifying nuclear material movements in terms of cask transfers and ensurance against diversion of nuclear material belong to the fundamental safeguards goals which have been achieved in the intermediate storage facilities by containment and surveillance techniques in unattended mode. (ii) Conditioning of spent fuel assemblies by separating the fuel rods from structural elements. Since the pilot conditioning facility in Gorleben has not yet come into operation, the underlying safeguards approach which focuses on safeguarding the key measurement points - the spent fuel related way in and out of the facility - has not been applied yet. (iii) Disposal in deep geological formations, but no decision has been made so far neither regarding the location of a geological repository nor regarding the safeguards approach for the disposal concept of spent fuel. The situation was complicated by a moratorium which suspended the underground exploration of the Gorleben salt dome as potential geological repository for spent fuel. The moratorium expires in October 2010. Nevertheless, considerable progress has been made in the development of disposal concepts. According to the basic, so-called POLLUX (registered) -concept spent fuel assemblies are to be conditioned after dry storage and reloaded into the POLLUX (registered) -cask
Projections of spent fuel to be discharged by the U.S. nuclear power industry

International Nuclear Information System (INIS)

Alexander, C.W.; Kee, C.W.; Croff, A.G.; Blomeke, J.O.

1977-10-01

Calculated properties of spent fuel projected to be discharged and accumulated by the U.S. nuclear power industry through the year 2031 A.D. are presented. The projections are based on installed nuclear capacities of 380 and 543 GW(e) in the year 2000 and 2030, respectively. They include compilations of the grams of the elements, curies of radioactivity, thermal decay power, photon and neutron emission rates, and radiotoxicities of the assemblies that are accumulated at a Spent Unreprocessed Fuel Facility (SURFF), allowing for delays of 5 and 10 years before shipment to SURFF
Projections of spent fuel to be discharged by the U. S. nuclear power industry

Energy Technology Data Exchange (ETDEWEB)

Alexander, C.W.; Kee, C.W.; Croff, A.G.; Blomeke, J.O.

1977-10-01

Calculated properties of spent fuel projected to be discharged and accumulated by the U.S. nuclear power industry through the year 2031 A.D. are presented. The projections are based on installed nuclear capacities of 380 and 543 GW(e) in the year 2000 and 2030, respectively. They include compilations of the grams of the elements, curies of radioactivity, thermal decay power, photon and neutron emission rates, and radiotoxicities of the assemblies that are accumulated at a Spent Unreprocessed Fuel Facility (SURFF), allowing for delays of 5 and 10 years before shipment to SURFF.
Waste form dissolution in bedded salt

International Nuclear Information System (INIS)

Kaufman, A.M.

1980-01-01

A model was devised for waste dissolution in bedded salt, a hydrologically tight medium. For a typical Spent UnReprocessed Fuel (SURF) emplacement, the dissolution rate wll be diffusion limited and will rise to a steady state value after t/sub eq/ approx. = 250 (1+(1-epsilon 0 ) K/sub D//epsilon 0 ) (years) epsilon 0 is the overpack porosity and K/sub d/ is the overpack sorption coefficient. The steady state dissolution rate itself is dominated by the solubility of UO 2 . Steady state rates between 5 x 10 -5 and .5 (g/year) are achievable by SURF emplacements in bedded salt without overpack, and rates between 5 x 10 -7 and 5 x 10 -3 (g/year) with an overpack having porosity of 10 -2

An analysis of the technical status of high level radioactive waste and spent fuel management systems

Science.gov (United States)

English, T.; Miller, C.; Bullard, E.; Campbell, R.; Chockie, A.; Divita, E.; Douthitt, C.; Edelson, E.; Lees, L.

1977-01-01

The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.
Effects of spent nuclear fuel aging on disposal requirements

International Nuclear Information System (INIS)

McKee, R.W.; Johnson, K.I.; Huber, H.D.; Bierschbach, M.C.

1991-10-01

This paper describes results of a study to analyze the waste management systems effects of extended spent fuel aging on spent fuel disposal requirements. The analysis considers additional spent fuel aging up to a maximum of 50 years relative to the currently planned 2010 repository startup in the United States. As part of the analysis, an equal energy disposition (EED) methodology was developed for determining allowable waste emplacement densities and waste container loading in a geologic repository. Results of this analysis indicate that substantial benefits of spent fuel aging will already have been achieved by a repository startup in 2010 (spent fuel average age will be 28 years). Even so, further significant aging benefits, in terms of reduced emplacement areas and mining requirements and reduced number of waste containers, will continue to accrue for at least another 50 years when the average spent fuel age would be 78 years, if the repository startup is further delayed
Logistics Modeling of Emplacement Rate and Duration of Operations for Generic Geologic Repository Concepts

Energy Technology Data Exchange (ETDEWEB)

Kalinina, Elena Arkadievna; Hardin, Ernest

2015-11-01

This study identified potential geologic repository concepts for disposal of spent nuclear fuel (SNF) and (2) evaluated the achievable repository waste emplacement rate and the time required to complete the disposal for these concepts. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel. The results of this study provide an important input for the rough-order-of-magnitude (ROM) disposal cost analysis. The disposal concepts cover three major categories of host geologic media: crystalline or hard rock, salt, and argillaceous rock. Four waste package sizes are considered: 4PWR/9BWR; 12PWR/21BWR; 21PWR/44BWR, and dual purpose canisters (DPCs). The DPC concepts assume that the existing canisters will be sealed into disposal overpacks for direct disposal. Each concept assumes one of the following emplacement power limits for either emplacement or repository closure: 1.7 kW; 2.2 kW; 5.5 kW; 10 kW; 11.5 kW, and 18 kW.
Logistics Modeling of Emplacement Rate and Duration of Operations for Generic Geologic Repository Concepts

International Nuclear Information System (INIS)

Kalinina, Elena Arkadievna; Hardin, Ernest

2015-01-01

This study identified potential geologic repository concepts for disposal of spent nuclear fuel (SNF) and (2) evaluated the achievable repository waste emplacement rate and the time required to complete the disposal for these concepts. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel. The results of this study provide an important input for the rough-order-of-magnitude (ROM) disposal cost analysis. The disposal concepts cover three major categories of host geologic media: crystalline or hard rock, salt, and argillaceous rock. Four waste package sizes are considered: 4PWR/9BWR; 12PWR/21BWR; 21PWR/44BWR, and dual purpose canisters (DPCs). The DPC concepts assume that the existing canisters will be sealed into disposal overpacks for direct disposal. Each concept assumes one of the following emplacement power limits for either emplacement or repository closure: 1.7 kW; 2.2 kW; 5.5 kW; 10 kW; 11.5 kW, and 18 kW.
Spent-Fuel Test - Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Executive summary of final results

International Nuclear Information System (INIS)

Patrick, W.C.

1986-01-01

This summary volume outlines results that are covered in more detail in the final report of the Spent-Fuel Test - Climate project. The project was conducted between 1978 and 1983 in the granitic Climax stock at the Nevada Test Site. Results indicate that spent fuel can be safely stored for periods of years in this host medium and that nuclear waste so emplaced can be safely retrieved. We also evaluated the effects of heat and radiation (alone and in combination) on emplacement canisters and the surrounding rock mass. Storage of the spent-fuel affected the surrounding rock mass in measurable ways, but did not threaten the stability or safety of the facility at any time
Operational considerations in drift emplacement of waste packages

International Nuclear Information System (INIS)

Benton, H.A.

1993-01-01

This paper discusses the operational considerations as well as the advantages and disadvantages of emplacing waste packages in drifts in a repository. The considerations apply particularly to the potential repository for spent nuclear fuel and high-level waste glass at Yucca Mountain, although most of the considerations and the advantages and disadvantages discussed in this paper do not necessarily represent the official views of the DOE or of the Management and Operations Contractor, since most of these considerations are still under active discussion and the final decisions will not be made for some time - perhaps years. This paper describes the issues, suggests some principles upon which decisions should be based, and states some of the most significant advantages and disadvantages of the emplacement modes, and the associated waste package types and thermal loadings
Comparison of measured and calculated radiation doses in granite around emplacement holes in the spent-fuel test: Climax, Nevada Test Site

International Nuclear Information System (INIS)

van Konynenburg, R.A.

1982-01-01

Lawrence Livermore National Laboratory (LLNL) has emplaced eleven spent nuclear-reactor fuel assemblies in the Climax granite at the Nevada Test Site as part of the DOE Nevada Nuclear-Waste Storage Investigations. One of our objectives is to study radiation effects on the rock. The neutron and gamma-ray doses to the rock have been determined by MORSE-L Monte Carlo calculations and measurements using optical absorption and thermoluminescence dosimeters and metal foils. We compare the results to date. Generally, good agreement is found in the spatial and time dependence of the doses, but some of the absolute dose results appear to differ by more than the expected uncertainties. Although the agreement is judged to be adequate for radiation effects studies, suggestions for improving the precision of the calculations and measurements are made
Three-dimensional thermal analysis of a baseline spent fuel repository

International Nuclear Information System (INIS)

Altenbach, T.J.; Lowry, W.E.

1980-01-01

A three-dimensional thermal analysis has been performed using finite difference techniques to determine the near-field response of a baseline spent fuel repository in a deep geologic salt medium. A baseline design incorporates previous thermal modeling experience and OWI recommendations for areal thermal loading in specifying the waste form properties, package details, and emplacement configuration. The base case in this thermal analysis considers one 10-year old PWR spent fuel assembly emplaced to yield a 36 kw/acre (8.9 w/m 2 ) loading. A unit cell model in an infinite array is used to simplify the problem and provide upper-bound temperatures. Boundary conditions are imposed which allow simulations to 1000 years. Variations studied include a comparison of ventilated and unventilated storage room conditions, emplacement packages with and without air gaps surrounding the canister, and room cool-down scenarios with ventilation following an unventilated state for retrieval purposes. At this low power level ventilating the emplacement room has an immediate cooling influence on the canister and effectively maintains the emplacement room floor near the temperature of the ventilating air. The annular gap separating the canister and sleeve causes the peak temperature of the canister surface to rise by 10 0 F (5.6 0 C) over that from a no gap case assuming perfect thermal contact. It was also shown that the time required for the emplacement room to cool down to 100 0 F (38 0 C) from an unventilated state ranged from 2 weeks to 6 months; when ventilation initiated after times of 5 years to 50 years, respectively. As the work was performed for the Nuclear Regulatory Commission, these results provide a significant addition to the regulatory data base for spent fuel performance in a geologic repository
Mechanical and thermomechanical calculations related to the storage of spent nuclear-fuel assemblies in granite

International Nuclear Information System (INIS)

Butkovich, T.R.

1980-05-01

A generic test of the geologic storage of spent-fuel assemblies is being made at Nevada Test Site. The spent-fuel assemblies were emplaced at a depth of 420 m (1370 ft) below the surface in a typical granite and will be retrieved at a later time. The early time, close-in thermal history of this type of repository is being simulated with spent-fuel and electrically heated canisters in a central drift, with auxiliary heaters in two parallel side drifts. Prior to emplacement of the spent-fuel canisters, preliminary calculations were made using a pair of existing finite-element codes, ADINA and ADINAT
Thermal/thermomechanical analyses for the room region with horizontal and vertial modes of emplacement

International Nuclear Information System (INIS)

1988-01-01

Extensive thermal/thermomechanical analyses of the Site Characterization Plan-Conceptual Design at the Deaf Smith county Site, Texas, have been carried out for the room region with horizontal and vertical modes of emplacement. The main purpose of this study is to make a good comparison between these two modes of emplacement in this region. Homogeneous and nonhomogeneous strata under isothermal or transient temperature conditions cases were considered in the analyses. Furthermore, various pillar widths for the vertical mode emplacement were also taken into consideration. Only spent fuel (SF) waste was considered in this study. Finite element method was used throughout the analyses. The thermal responses were evaluated using SPECTROM-41 while the thermomechanical responses were calculated using SPECTROM-32. Thermal and thermomechanical comparisons between the two modes of emplacement for various cases were presented in this paper
Advanced three-dimensional thermal modeling of a baseline spent fuel repository

International Nuclear Information System (INIS)

Altenbach, T.J.; Lowry, W.E.

1980-01-01

A three-dimensional thermal analysis using finite difference techniques was performed to determine the near-field response of a baseline spent fuel repository in a deep geologic salt medium. A baseline design incorporates previous thermal modeling experience and OWI recommendations for areal thermal loading in specifying the waste form properties, package details, and emplacement configuration. The base case in this thermal analysis considers one 10-year old PWR spent fuel assembly emplaced to yield a 36 kW/acre (8.9 W/m 2 ) loading. A unit cell model in an infinite array is used to simplify the problem and provide upper-bound temperatures. Boundary conditions are imposed which allow simulations to 1000 years. Variations studied include a comparison of ventilated and unventilated storage room conditions, emplacement packages with and without air gaps surrounding the canister, and room cool-down scenarios with ventilation following an unventilated state for retrieval purposes. It was found that at this low-power level, ventilating the emplacement room has an immediate cooling influence on the canister and effectively maintains the emplacement room floor near the temperature of the ventilating air
Lessons learned in demonstration projects regarding operational safety during final disposal of vitrified waste and spent fuel

International Nuclear Information System (INIS)

Filbert, Wolfgang; Herold, Philipp

2015-01-01

The paper summarizes the lessons learned in demonstration projects regarding operational safety during the final disposal of vitrified waste and spent fuel. The three demonstration projects for the direct disposal of vitrified waste and spent fuel are described. The first two demonstration projects concern the shaft transport of heavy payloads of up to 85 t and the emplacement operations in the mine. The third demonstration project concerns the borehole emplacement operation. Finally, open issues for the next steps up to licensing of the emplacement and disposal systems are summarized.
Disposal Of Spent Fuel In Salt Using Borehole Technology: BSK 3 Concept

Energy Technology Data Exchange (ETDEWEB)

Fopp, Stefan; Graf, Reinhold [GNS Gesellschaft fuer Nuklear-Service mbH, Hollestrasse 7A, D-45127 Essen (Germany); Filbert, Wolfgang [DBE TECHNOLOGY GmbH, Eschenstrasse 55, D-31224 Peine (Germany)

2008-07-01

The BSK 3 concept was developed for the direct disposal of spent fuel in rock salt. It is based on the conditioning of fuel assemblies and inserting fuel rods into a steel canister which can be placed in vertical boreholes. The BSK 3 canister is suitable for spent fuel rods from 3 PWR or 9 BWR fuel assemblies. The emplacement system developed for the handling and disposal of BSK 3 canisters comprises a transfer cask which provides appropriate shielding during the transport and emplacement process, a transport cart, and an emplacement device. Using the emplacement device the transfer cask will be positioned onto the top of the borehole lock. The presentation describes the development and the design of the transfer cask and the borehole lock. A technically feasible and safe design for the transfer cask and the borehole lock was found regarding the existing safety requirements for radiation shielding, heat dissipation and handling procedure. (authors)
German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

Directory of Open Access Journals (Sweden)

A. Schwenk-Ferrero

2013-01-01

Full Text Available Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104 to 106 years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.
Spent fuel test. Climax data acquisition system integration report

International Nuclear Information System (INIS)

Nyholm, R.A.; Brough, W.G.; Rector, N.L.

1982-06-01

The Spent Fuel Test - Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granitic rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. This multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system collects and processes data from more than 900 analog instruments. This report documents the design and functions of the hardware and software elements of the Data Acquisition System and describes the supporting facilities which include environmental enclosures, heating/air-conditioning/humidity systems, power distribution systems, fire suppression systems, remote terminal stations, telephone/modem communications, and workshop areas. 9 figures
Spent Fuel Test - Climax data acquisition system operations manual

International Nuclear Information System (INIS)

Nyholm, R.A.

1983-01-01

The Spent Fuel Test-Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granite rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the US Department of Energy Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. The multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system (DAS) collects and processes data from more than 900 analog instruments. This report documents the software element of the LLNL developed SFT-C Data Acquisition System. It defines the operating system and hardware interface configurations, the special applications software and data structures, and support software
Review of the KBS II plan for handling and final storage of unreprocessed spent nuclear fuel

International Nuclear Information System (INIS)

1980-01-01

The Swedish utilities programme for disposal of spent nuclear fuel elements (KBS II) is summarized. Comments and criticism to the programme are given by experts from several foreign or international institutions. (L.E.)
Conceptual designs for waste packages for horizontal or vertical emplacement in a repository in salt for reference in the site characterization plan

International Nuclear Information System (INIS)

1987-06-01

This report includes the options of horizontal and vertical emplacement, the addition of a phased repository, an additional waste form (intact spent fuel), revised geotechnical data appropriate for the Deaf Smith County site, new corrosion data for the container, and new repository design data. The waste package consists of waste form and canister within a thick-walled, low-carbon steel container surrounded by packing. The container is a hollow cylinder with a flat head welded to each end. The design concepts for the waste container or vertical and horizontal emplacement are identical. This report discusses the results of analyses of aspects of the reference waste package concept needing changes because of new data and information believed applicable to the Deaf Smith County site. Included are waste package conceptual designs or (1) the reference defense high-level waste form from the Savannah River Plant; (2) intact spent fuel with our pressurized-water-reactor or nine boiling-water-reactor assemblies per package for emplacement during Phase 1 of repository operation; and (3) spent fuel which has been disassembled and consolidated into a segmented cylindrical canister with rods from either 12 pressurized-water-reactor or 30 boiling-water-reactor assemblies per package for emplacement during Phase 2. 30 refs., 61 figs., 30 tabs
The Swedish Concept for Disposal of Spent Nuclear Fuel: Differences Between Vertical and Horizontal Waste Canister Emplacement

International Nuclear Information System (INIS)

Bennett, D.G.; Hicks, T.W.

2005-10-01

The Swedish Nuclear Power Inspectorate (SKI) is preparing for the review of licence applications related to the disposal of spent nuclear fuel. The Swedish Nuclear Fuel and Waste Management Company (SKB) refers to its proposals for the disposal of spent nuclear fuel as the KBS-3 concept. In the KBS-3 concept, SKB plans that, after 30 to 40 years of interim storage, spent fuel will be disposed of at a depth of about 500 m in crystalline bedrock, surrounded by a system of engineered barriers. The principle barrier to radionuclide release is a cylindrical copper canister. Within the copper canister, the spent fuel is supported by a cast iron insert. Outside the copper canister is a layer of bentonite clay, known as the buffer, which is designed to provide mechanical protection for the canisters and to limit the access of groundwater and corrosive substances to their surfaces. The bentonite buffer is also designed to sorb radionuclides released from the canisters, and to filter any colloids that may form within the waste. SKB is expected to base its forthcoming licence applications on a repository design in which the waste canisters are emplaced in vertical boreholes (KBS-3V). However, SKB has also indicated that it might be possible and, in some respects, beneficial to dispose of the waste canisters in horizontal tunnels (KBS-3H). There are many similarities between the KBS-3V and KBS-3H designs. There are, however, uncertainties associated with both of the designs and, when compared, both possess relative advantages and disadvantages. SKB has identified many of the key factors that will determine the evolution of a KBS-3H repository and has plans for research and development work in many of the areas where the differences between the KBS-3V and KBS-3H designs mean that they could be significant in terms of repository performance. With respect to the KBS-3H design, key technical issues are associated with: 1. The accuracy of deposition drift construction. 2. Water
Spent fuel management in Canada

International Nuclear Information System (INIS)

Pattantyus, P.

1998-01-01

The current status of the Canadian Spent Fuel Management is described. This includes wet and dry interim storage, transportation issues and future plans regarding final disposal based on deep underground emplacement in stable granite rock. Extension of wet interim storage facilities is not planned, as dry storage technologies have found wide acceptance. (author)

Thermal calculations for the design, construction, operation, and evaluation of the Spent Fuel Test - Climax, Nevada Test Site

International Nuclear Information System (INIS)

Montan, D.N.; Patrick, W.C.

1981-01-01

The Spent Fuel Test-Climax (SFT-C) is a test of retrievable deep geologic storage of commercially generated spent nuclear reactor fuel in granitic rock. Eleven spent fuel assemblies, together with six electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the US Department of Energy Nevada Test Site. On June 2, 1978 LLNL secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. This report documents a series of thermal calculations that were performed in support of the SFT-C. Early calculations employed analytical solutions to address such design and construction issues as drift layout and emplacement hole spacings. Operational aspects of the test required more detailed numerical solutions dealing with ventilation and guard-heater power levels. The final set of calculations presented here provides temperature histories throughout the test facility for evaluation of the response of the SFT-C and for comparison of calculations with acquired data. This final set of calculations employs the as-built test geometry and best-available material properties
Emplacement of small and large buffer blocks

International Nuclear Information System (INIS)

Saari, H.; Nikula, M.; Suikki, M.

2010-05-01

The report describes emplacement of a buffer structure encircling a spent fuel canister to be deposited in a vertical hole. The report deals with installability of various size blocks and with an emplacement gear, as well as evaluates the achieved quality of emplacement and the time needed for installing the buffer. Two block assembly of unequal size were chosen for examination. A first option involved small blocks, the use of which resulted in a buffer structure consisting of small sector blocks 200 mm in height. A second option involved large blocks, resulting in a buffer structure which consists of eight blocks. In these tests, the material chosen for both block options was concrete instead of bentonite. The emplacement test was a three-phase process. A first phase included stacking a two meter high buffer structure with small blocks for ensuring the operation of test equipment and blocks. A second phase included installing buffer structures with both block options to a height matching that of a canister-encircling cylindrical component. A third phase included testing also the installability of blocks to be placed above the canister by using small blocks. In emplacement tests, special attention was paid to the installability of blocks as well as to the time required for emplacement. Lifters for both blocks worked well. Due to the mass to be lifted, the lifter for large blocks had a more heavy-duty frame structure (and other lifting gear). The employed lifters were suspended in the tests on a single steel wire rope. Stacking was managed with both block sizes at adequate precision and stacked-up towers were steady. The stacking of large blocks was considerably faster. Therefore it is probably that the overall handling of the large blocks will be more convenient at a final disposal site. From the standpoint of reliability in lifting, the small blocks were safer to install above the canister. In large blocks, there are strict shape-related requirements which are
Technical concept for a test of geologic storage of spent reactor fuel in the climax granite, Nevada Test Site

International Nuclear Information System (INIS)

Ramspott, L.D.; Ballou, L.B.; Carlson, R.C.; Montan, D.N.; Butkovich, T.R.; Duncan, J.E.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Mayr, M.C.

1979-01-01

We plan to emplace spent fuel assemblies from an operating commercial nuclear reactor in the Climax granite at the US Department of Energy's Nevada Test Site. In this generic test, 11 canisters of spent fuel will be emplaced with 6 electrical simulator canisters in a storage drift 420 m below in surface and their effects compared. Two adjacent drifts will contain electrical heaters, operated to simulate the temperature-stress-displacement fields of a large repository. We describe the test objectives, the technical issues, the site, the preoperational measurement program, thermal and mechanical response calculations, the characteristics of the spent fuel, the field instrumentation and data-acquisition systems, and the system for handling the spent fuel
Technical concept for test of geologic storage of spent reactor fuel in the Climax granite, Nevada Test Site

International Nuclear Information System (INIS)

Ramspott, L.D.; Ballou, L.B.; Carlson, R.C.; Montan, D.N.; Butkovich, T.R.; Duncan, J.E.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Mayr, M.C.

1979-01-01

The Spent Fuel Test in the Climax granite at the Nevada Test Site is a generic test in which spent fuel assemblies from an operating commercial nuclear reactor are emplaced at, and retrieved from, a plausible waste repository depth in a typical granite. Eleven canisters of spent fuel are emplaced in a storage drift 420 m below the surface along with six electrical simulator canisters. Two adjacent drifts contain electrical heaters which are operated so as to simulate the initial five years of the temperature-stress-displacement fields of a large repository. The site is described, and the pre-operational measurement program and characteristics of the spent fuel are given. Both thermal and mechanical response calculations are summarized. The field instrumentation and data acquisition systems are described, as well as the system for handling the spent fuel
Global Spent Fuel Logistics Systems Study (GSFLS). Volume 4. Pacific basin spent fuel logistics system

International Nuclear Information System (INIS)

1978-06-01

This report summarizes the conceptual framework for a Pacific Basin Spent Fuel Logistics System (PBSFLS); and preliminarily describes programatic steps that might be taken to implement such a system. The PBSFLS concept is described in terms of its technical and institutional components. The preferred PBSFLS concept embodies the rationale of emplacing a fuel cycle system which can adjust to the technical and institutional non-proliferation solutions as they are developed and accepted by nations. The concept is structured on the basis of initially implementing a regional spent fuel storage and transportation system which can technically and institutionally accommodate downstream needs for energy recovery and long-term waste management solutions
Near-field heat transfer at the spent fuel test-climax: a comparison of measurements and calculations

International Nuclear Information System (INIS)

Patrick, W.C.; Montan, D.N.; Ballou, L.B.

1981-01-01

The Spent Fuel Test in the Climax granitic stock at the DOE Nevada Test Site is a test of the feasibility of storage and retrieval of spent nuclear reactor fuel in a deep geologic environment. Eleven spent fuel elements, together with six thermally identical electrical resistance heaters and 20 peripheral guard heaters, are emplaced 420 m below surface in a three-drift test array. This array was designed to simulate the near-field effects of thousands of canisters of nuclear waste and to evaluate the effects of heat alone, and heat plus ionizing radiation on the rock. Thermal calculations and measurements are conducted to determine thermal transport from the spent fuel and electrical resistance heaters. Calculations associated with the as-built Spent Fuel Test geometry and thermal source histories are presented and compared with thermocouple measurements made throughout the test array. Comparisons in space begin at the spent fuel canister and include the first few metres outside the test array. Comparisons in time begin at emplacement and progress through the first year of thermal loading in this multi-year test
Logistics models for the transportation of radioactive waste and spent fuel

International Nuclear Information System (INIS)

Joy, D.S.; Holcomb, B.D.

1978-03-01

Mathematical modeling of the logistics of waste shipment is an effective way to provide input to program planning and long-range waste management. Several logistics models have been developed for use in parametric studies, contingency planning, and management of transportation networks. These models allow the determination of shipping schedules, optimal routes, probable transportation modes, minimal costs, minimal personnel exposure, minimal transportation equipment, etc. Such information will permit OWI to specify waste-receiving rates at various repositories in order to balance work loads, evaluate surge capacity requirements, and estimate projected shipping cask fleets. The programs are tailored to utilize information on the types of wastes being received, location of repositories and waste-generating facilities, shipping distances, time required for a given shipment, availability of equipment, above-ground storage capabilities and locations, projected waste throughput rates, etc. Two basic models have been developed. The Low-Level Waste Model evaluates the optimal transportation policy for shipping waste directly from the source to a final destination without any intermediate stops. The Spent Fuel Logistics Model evaluates the optimal transportation policy for shipping unreprocessed spent fuel from nuclear power plants (1) indirectly, that is, to an Away-From-Reactor (AFR) storage facility, with subsequent transhipment to a repository, or (2) directly to a repository
Uncanistered Spent Nuclear fuel Disposal Container System Description Document

International Nuclear Information System (INIS)

Pettit, N. E.

2001-01-01

The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident
Equipment designs for the spent LWR fuel dry storage demonstration

International Nuclear Information System (INIS)

Steffen, R.J.; Kurasch, D.H.; Hardin, R.T.; Schmitten, P.F.

1980-01-01

In conjunction with the Spent Fuel Handling and Packaging Program (SFHPP) equipment has been designed, fabricated and successfully utilized to demonstrate the packaging and interim dry storage of spent LWR fuel. Surface and near surface storage configurations containing PWR fuel assemblies are currently on test and generating baseline data. Specific areas of hardware design focused upon include storage cell components and the support related equipment associated with encapsulation, leak testing, lag storage, and emplacement operations
Preliminary estimates of the charge for spent-fuel storage and disposal services

International Nuclear Information System (INIS)

1978-07-01

The purpose of this report is to stimulate discussion among a wide range of interested parties concerning a one-time charge by the U.S. Government for disposal, or interim storage and disposal, of spent unreprocessed nuclear fuel. The report contains a set of estimates of the charge based on current cost figures and a variety of demand, logistical, institutional, and cost overrun assumptions. The services are to be offered to domestic utilities by the U.S. Government in connection with the spent fuel policy approved by the President and announced by the Department of Energy (DOE) on October 18, 1977. This policy is a direct result of the indefinite deferral of all commercial reprocessing of spent fuel announced by President Carter on April 7, 1977. The services will also be offered to foreign governments on a limited basis in cases where this action would contribute to U.S. goals for nonproliferation of nuclear weapons. The report does not establish new policy and it does not commit DOE to any specific program, schedule or charge. No scenario or case is to be considered most important, no methodology is to be considered definitive, and no charge is to be considered most likely or to represent a proposed charge. The report describes basic principles and methodologies for calculating the charge and highlights primary cost centers. Current estimates of program and facility costs are used. Various aspects of the DOE Spent Fuel Storage Program are brought into focus through this analysis. Interested parties should find these assessment criteria helpful for their planning and useful in discussions concerning the program
Geomechanics of the Spent Fuel Test: Climax

International Nuclear Information System (INIS)

Wilder, D.G.; Yow, J.L. Jr.

1987-07-01

Three years of geomechanical measurements were made at the Spent Fuel Test-Climax (SFT-C) 1400 feet underground in fractured granitic rock. Heating of the rock mass resulted from emplacement of spent fuel as well as the heating by electrical heaters. Cooldown of the rock occurred after the spent fuel was removed and the heaters were turned off. The measurements program examines both gross and localized responses of the rock mass to thermal loading, to evaluate the thermomechanical response of sheared and fractured rock with that of relatively unfractured rock, to compare the magnitudes of displacements during mining with those induced by extensive heating of the rock mass, and to check assumptions regarding symmetry and damaged zones made in numerical modeling of the SFT-C. 28 refs., 113 figs., 10 tabs
Disposal of spent fuel from German nuclear power plants - 16028

International Nuclear Information System (INIS)

Graf, Reinhold; Brammer, Klaus-Juergen; Filbert, Wolfgang; Bollingerfehr, Wilhelm

2009-01-01

The 'direct disposal of spent fuel' as a part of the current German reference concept was developed as an alternative to spent fuel reprocessing and vitrified HLW disposal. The technical facilities necessary for the implementation of this part of the reference concept, the so called POLLUX R concept, i.e. interim storage buildings for casks containing spent fuel, a pilot conditioning facility, and a special cask 'POLLUX' for final disposal have been built. With view to a geological salt formation all handling procedures for the direct disposal of spent fuel were tested aboveground in full-scale test facilities. To optimise the reference concept, all operational steps have been reviewed for possible improvements. The two additional concepts for the direct disposal of SF are the BSK 3 concept and the DIREGT concept. Both concepts rely on borehole emplacement technology, vertical boreholes for the BSK 3 concept und horizontal boreholes for the DIREGT concept. Supported by the EU and the German Federal Ministry of Economics and Technology (BMWi), DBE TECHNOLOGY built an aboveground full-scale test facility to simulate all relevant handling procedures for the BSK 3 disposal concept. GNS (Company for Nuclear Service), representing the German utilities, provided the main components and its know-how concerning cask design and manufacturing. The test program was concluded recently after more than 1.000 emplacement operations had been performed successfully. The BSK 3 emplacement system in total comprises an emplacement device, a borehole lock, a transport cart, a transfer cask which will shuttle between the aboveground conditioning facility and the underground repository, and the BSK 3 canister itself, designed to contain the fuel rods of three PWR-fuel assemblies with a total of about 1.6 tHM. The BSK 3 concept simplifies the operation of the repository because the handling procedures and techniques can also be applied for the disposal of reprocessing residues. In addition
Spent Fuel Test-Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Final report

International Nuclear Information System (INIS)

Patrick, W.C.

1986-01-01

In the Climax stock granite on the Nevada Test Site, eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized. When test data indicated that the test objectives were met during the 3-year storage phase, the spent-fuel canisters were retrieved and the thermal sources were de-energized. The project demonstrated the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner. In addition to emplacement and retrieval operations, three exchanges of spent-fuel assemblies between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. The test led to development of a technical measurements program. To meet these objectives, nearly 1000 instruments and a computer-based data acquisition system were deployed. Geotechnical, seismological, and test status data were recorded on a continuing basis for the three-year storage phase and six-month monitored cool-down of the test. This report summarizes the engineering and scientific endeavors which led to successful design and execution of the test. The design, fabrication, and construction of all facilities and handling systems are discussed, in the context of test objectives and a safety assessment. The discussion progresses from site characterization and experiment design through data acquisition and analysis of test data in the context of design calculations. 117 refs., 52 figs., 81 tabs
Spent Fuel Test-Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Final report

Energy Technology Data Exchange (ETDEWEB)

Patrick, W.C. (comp.)

1986-03-30

In the Climax stock granite on the Nevada Test Site, eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized. When test data indicated that the test objectives were met during the 3-year storage phase, the spent-fuel canisters were retrieved and the thermal sources were de-energized. The project demonstrated the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner. In addition to emplacement and retrieval operations, three exchanges of spent-fuel assemblies between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. The test led to development of a technical measurements program. To meet these objectives, nearly 1000 instruments and a computer-based data acquisition system were deployed. Geotechnical, seismological, and test status data were recorded on a continuing basis for the three-year storage phase and six-month monitored cool-down of the test. This report summarizes the engineering and scientific endeavors which led to successful design and execution of the test. The design, fabrication, and construction of all facilities and handling systems are discussed, in the context of test objectives and a safety assessment. The discussion progresses from site characterization and experiment design through data acquisition and analysis of test data in the context of design calculations. 117 refs., 52 figs., 81 tabs.
NWTS conceptual reference repository description (CRRD). Volume 1. Summary

International Nuclear Information System (INIS)

1981-05-01

This report is a Conceptual Reference Repository Description (CRRD) for the terminal storage of spent, unreprocessed fuel assemblies and low-level transuranic waste. It is based on the following three previously prepared conceptual design reports: National Waste Terminal Storage in a Bedded Salt Formation for Spent Unreprocessed Fuel (NWTS-R2), prepared by Kaiser Engineers; National Waste Terminal Storage Repository Number 1 (NWTS-R1), prepared by Stearns-Roger; Spent Fuel Receiving and Packaging Facility Conceptual Design, prepared by Rockwell Hanford Operations and Kaiser Engineers. The CRRD is not itself a conceptual design report, but combines the key features of the above three reports into a description of a reference repository. This description is to be used as a basis for the preparation of a preliminary information report for the repositories. The CRRD discusses from a conceptual engineering standpoint the structures, systems, equipment, and operations necessary to (1) receive unreprocessed spent fuel assemblies via standard shipping casks, (2) offload these assemblies, (3) place the assemblies in canisters, (4) transport the canisters to underground storage locations in the salt dome, and (5) place these canisters in terminal storage. The CRRD also elaborates on the concepts for the retrieval and recovery of the spent fuel after burial; describes the development of the shafts and the underground areas,as well as the supporting operational utility and administrative features of the repository; and discusses the impacts on the reference repository of (1) changes in the waste package concept, (2) dry receiving and dry temporary storage of waste, (3) terminal storage of reprocessing waste, and (4) alternative underground development schemes
Elements of transport and emplacement system

International Nuclear Information System (INIS)

1981-02-01

This report, undertaken to review proposals for transport, handling and emplacement of high-level radioactive wastes in an underground repository, appropriate to the U.K. context, falls under the headings: basic design concepts; waste block size and configuration; self-shielded or partially shielded blocks; concept of emplacement in long boreholes; concept of emplacement in short boreholes; concept of emplacement in tunnels; methods of emplacement; stages of disposal; repository access by adit, incline or shaft; handling techniques within repository; conventional and radiological safety; costs; areas for further research and development. (U.K.)
Hydrothermal interactions of cesium and strontium phases from spent unreprocessed fuel with basalt phases and basalts

International Nuclear Information System (INIS)

Komarneni, S.; Scheetz, B.E.; McCarthy, G.J.; Coons, W.E.

1980-03-01

This investigation is a segment of an extensive research program aimed at investigating the feasibility of long-term, subsurface storage of commercial nuclear waste. Specifically, it is anticipated that the waste will be housed in a repository mined from the basalt formations which lie beneath the Hanford Site. The elements monitored during the present experiments were Cs and Sr. These two elements represent significant biohazards if released from a repository and are the major heat producing radionuclides present in commercial radioactive waste. Several Cs phases and/or solutions were reacted with either isolated basalt phases or bulk-rock basalt, and the resulting solids and solutions were analyzed. The hydrothermal reactivity of SrZrO 3 , which is believed to be a probable host for Sr in SFE was investigated. While so far no evidence exists which indicates that Sr is present in a water soluble phase in spent fuel elements (SFE), detailed investigation of a potential hazard is warranted. This investigation has determined that some Cs compounds likely to be stable components of spent fuel (i.e., CsOH, Cs 2 MoO 4 , Cs 2 U 2 O 7 ) have significant hydrothermal solubilities. These solubilities are greatly decreased in the presence of basalt and/or basalt minerals. The decrease in the amount of Cs in solution results from reactions which form pollucite and/or CsAlSiO 4 , with the production of pollucite exceeding that of CsAlSiO 4 . Dissolution of β-Cs 2 U 2 O 7 implies solubilizing a uranium species to an undetermined extent. The production of schoepite (UO 3 .3H 2 O) during some experiments containing basalt phases, indicates a tendency to oxidize U 4+ to U 6+ . When diopside (nominally CaMgSi 2 O 6 ) and β-Cs 2 U 2 O 7 were hydrothermally reacted, at 300 0 C both UO 2 and UO 3 .3H 2 O were produced. Experiments on SrZrO 3 show it to be an unreactive phase
Aluminium oxide as an encapsulation material for unreprocessed nuclear fuel waste - evaluation from the viewpoint of corrosion

International Nuclear Information System (INIS)

1980-03-01

The Nuclear Fuel Safety Project (KBS) has proposed that spent unreprocessed nuclear fuel shall be disposed of by encapsulation in canisters of high-purity alumina sintered under isostatic pressure. The canisters will have a wall thickness of 100 mm and are to be placed in vertical boreholes extending from horizontal tunnels 500 m below ground in igneous rock. In each borehole one canister is deposited embedded in a quartz sand/bentonite buffer. An expert group of 10 Swedish specialists has arrived at the following conclusions. The alumina is not thermodynamically stable in water. In pure water hydration will occur, below 100degC leading to the formation of either Al(OH) 3 in the amorphous state or crystalline gibbsite (Al 2 O 3 x 3H 2 O). Corrosion may take place by slow dissolution or flaking off of a surface layer. Various immersion tests showed that the corrosion rate will be less than 0.1 μm/year, probably one or two powers of ten lower. If the alumina canister in the storage has sufficiently large surface defects and is under sufficiently high mechanical tension the defects may grow slowly into propagating cracks, ultimately leading to fracture, so-called delayed fracture. On the basis of results from fracture mechanical studies and after introduction of safety factors with respect to possible unknown features of the delayed fracture it was judged possible to eliminate the risk of delayed fracture if the canisters pass the following production control: - Proof testing at 150 MN/m 2 , using acoustic emission technique to ensure that crack growth does not occur during the unstressing cycle. - Surface acoustic wave examination with respect to surface inclusions, canisters with inclusions larger than 100 μm within a 100 μm deep surface zone being rejected. Canisters which pass the production control mentioned are estimated to have a life of hundreds of thousands of years. (author)
Emplacement engineering

Energy Technology Data Exchange (ETDEWEB)

Hill, Ernest E [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

1970-05-01

Emplacement Engineering can be defined as that portion of a nuclear explosive project that is concerned with the emplacement of the explosive. This definition would then include virtually everything except the design and fabrication of the explosive and the post-shot-effects program. For future commercial application, the post-shot-effects program will essentially disappear. This emplacement portion of a nuclear explosive project constitutes a large fraction of the total project cost, but it has largely been overshadowed by the explosive and explosive-effects portions. As we move into commercial applications. Emplacement Engineering must receive more attention from both industry and government. To place emplacement costs in their proper relationship with total projects costs, we have performed a study of commercial underground nuclear explosive applications such as gas stimulation. Although there are many intangibles in such a study, we have been able to at least obtain some feel for the relative fractional costs of the non-explosive costs compared with the explosive costs. This study involved estimating the cost elements for applications using a single explosive at 5,000 ft, 10,000 ft, and 15,000 ft. For each depth, the cost estimates were made for a range of emplacement hole and explosive diameters. Results of these estimates for explosive-related costs, hole-related costs, and total costs are shown for the three depths. Note that the explosive package outside diameter is assumed as 2 inches less than the hole (or casing) inside diameter for all cases. For the 5,000-ft application the explosive-related costs dominate, and of particular importance is the indicated diameter for minimum total cost which occurs at approximately a 17.5-in. hole (15.5-in. explosive). The hole-related costs are in 'the same range as the explosive-related costs for the 10,000-ft application. For this case, the minimum total cost occurs at approximately a 14-in. hole (12-in. explosive
Excavation and drilling at a spent-fuel test facility in granitic rock

Energy Technology Data Exchange (ETDEWEB)

Patrick, W.C.; Mayr, M.C.

1981-10-01

Funding for a project to test the feasibility of safe and reliable storage and retrieval of spent fuel from a commercial nuclear reactor was approved by the Department of Energy on June 2, 1978. By May 28, 1980, 11 spent-fuel assemblies had been emplaced 420 m below the surface in the Climax granitic stock at the Nevada Test Site. Design and construction of the Spent Fuel Test-Climax, including fuel emplacement, had taken less than two years, at a total cost of $18.4 million. Construction activities were preceded by geologic exploration using four cored holes and existing underground workings. The sinking of a 0.76-m-diam shaft to the 420-m level initiated construction at the site. Effective rates of sinking varied from 0.16 m/h with a rotary tricone drill to 0.5 m/h with a hammer drill. Underground excavation included a central canister-storage drift 4.6 x 6.1 x 64 m long, two parallel 3.4 x 3.4-m heater drifts, and a tail drift. About 6700 m{sup 3} were excavated at an average rate of 2 m{sup 3}/h, and 178 cored holes, with diameters from 38 to 152 mm, were drilled. A total length of nearly 1100 m was drilled at rates ranging from 0.4 m/h to 1 m/h, depending on hole size and drilling equipment. Eighteen 610-mm-diam canister emplacement holes were hammer-drilled at an average rate of 1.4 m/h. The use of the critical path method, integrated contractors, and close cooperation between project participants facilitated completion of the project on schedule.

Excavation and drilling at a spent-fuel test facility in granitic rock

International Nuclear Information System (INIS)

Patrick, W.C.; Mayr, M.C.

1981-10-01

Funding for a project to test the feasibility of safe and reliable storage and retrieval of spent fuel from a commercial nuclear reactor was approved by the Department of Energy on June 2, 1978. By May 28, 1980, 11 spent-fuel assemblies had been emplaced 420 m below the surface in the Climax granitic stock at the Nevada Test Site. Design and construction of the Spent Fuel Test-Climax, including fuel emplacement, had taken less than two years, at a total cost of $18.4 million. Construction activities were preceded by geologic exploration using four cored holes and existing underground workings. The sinking of a 0.76-m-diam shaft to the 420-m level initiated construction at the site. Effective rates of sinking varied from 0.16 m/h with a rotary tricone drill to 0.5 m/h with a hammer drill. Underground excavation included a central canister-storage drift 4.6 x 6.1 x 64 m long, two parallel 3.4 x 3.4-m heater drifts, and a tail drift. About 6700 m 3 were excavated at an average rate of 2 m 3 /h, and 178 cored holes, with diameters from 38 to 152 mm, were drilled. A total length of nearly 1100 m was drilled at rates ranging from 0.4 m/h to 1 m/h, depending on hole size and drilling equipment. Eighteen 610-mm-diam canister emplacement holes were hammer-drilled at an average rate of 1.4 m/h. The use of the critical path method, integrated contractors, and close cooperation between project participants facilitated completion of the project on schedule
Hydrothermal interactions of cesium and strontium phases from spent unreprocessed fuel with basalt phases and basalts

Energy Technology Data Exchange (ETDEWEB)

Komarneni, S.; Scheetz, B.E.; McCarthy, G.J.; Coons, W.E.

1980-03-01

This investigation is a segment of an extensive research program aimed at investigating the feasibility of long-term, subsurface storage of commercial nuclear waste. Specifically, it is anticipated that the waste will be housed in a repository mined from the basalt formations which lie beneath the Hanford Site. The elements monitored during the present experiments were Cs and Sr. These two elements represent significant biohazards if released from a repository and are the major heat producing radionuclides present in commercial radioactive waste. Several Cs phases and/or solutions were reacted with either isolated basalt phases or bulk-rock basalt, and the resulting solids and solutions were analyzed. The hydrothermal reactivity of SrZrO/sub 3/, which is believed to be a probable host for Sr in SFE was investigated. While so far no evidence exists which indicates that Sr is present in a water soluble phase in spent fuel elements (SFE), detailed investigation of a potential hazard is warranted. This investigation has determined that some Cs compounds likely to be stable components of spent fuel (i.e., CsOH, Cs/sub 2/MoO/sub 4/, Cs/sub 2/U/sub 2/O/sub 7/) have significant hydrothermal solubilities. These solubilities are greatly decreased in the presence of basalt and/or basalt minerals. The decrease in the amount of Cs in solution results from reactions which form pollucite and/or CsAlSiO/sub 4/, with the production of pollucite exceeding that of CsAlSiO/sub 4/. Dissolution of ..beta..-Cs/sub 2/U/sub 2/O/sub 7/ implies solubilizing a uranium species to an undetermined extent. The production of schoepite (UO/sub 3/.3H/sub 2/O) during some experiments containing basalt phases, indicates a tendency to oxidize U/sup 4 +/ to U/sup 6 +/. When diopside (nominally CaMgSi/sub 2/O/sub 6/) and ..beta..-Cs/sub 2/U/sub 2/O/sub 7/ were hydrothermally reacted, at 300/sup 0/C both UO/sub 2/ and UO/sub 3/.3H/sub 2/O were produced. Results of experiments on SrZrO/sub 3/ show it to be
Emplacement of Columbia River flood basalt

Science.gov (United States)

Reidel, Stephen P.

1998-11-01

Evidence is examined for the emplacement of the Umatilla, Wilbur Creek, and the Asotin Members of Columbia River Basalt Group. These flows erupted in the eastern part of the Columbia Plateau during the waning phases of volcanism. The Umatilla Member consists of two flows in the Lewiston basin area and southwestern Columbia Plateau. These flows mixed to form one flow in the central Columbia Plateau. The composition of the younger flow is preserved in the center and the composition of the older flow is at the top and bottom. There is a complete gradation between the two. Flows of the Wilbur Creek and Asotin Members erupted individually in the eastern Columbia Plateau and also mixed together in the central Columbia Plateau. Comparison of the emplacement patterns to intraflow structures and textures of the flows suggests that very little time elapsed between eruptions. In addition, the amount of crust that formed on the earlier flows prior to mixing also suggests rapid emplacement. Calculations of volumetric flow rates through constrictions in channels suggest emplacement times of weeks to months under fast laminar flow for all three members. A new model for the emplacement of Columbia River Basalt Group flows is proposed that suggests rapid eruption and emplacement for the main part of the flow and slower emplacement along the margins as the of the flow margin expands.
TRANSPORT AND EMPLACEMENT EQUIPMENT DESCRIPTIONS

International Nuclear Information System (INIS)

1997-01-01

The objective and the scope of this document are to list and briefly describe the major mobile equipment necessary for waste package (WP) Transport and Emplacement in the proposed subsurface nuclear waste repository at Yucca Mountain. Primary performance characteristics and some specialized design features of the equipment are explained and summarized in the individual subsections of this document. The Transport and Emplacement equipment described in this document consists of the following: (1) WP Transporter; (2) Reusable Rail Car; (3) Emplacement Gantry; (4) Gantry Carrier; and (5) Transport Locomotive
Ground Control for Emplacement Drifts for LA

International Nuclear Information System (INIS)

Y. Sun

2004-01-01

The purpose of this calculation is to analyze the stability of repository emplacement drifts during the preclosure period, and to provide a final ground support method for emplacement drifts for the License Application (LA). The scope of the work includes determination of input parameter values and loads, selection of appropriate process and methods for the calculation, application of selected methods, such as empirical or analytical, to the calculation, development and execution of numerical models, and evaluation of results. Results from this calculation are limited to use for design of the emplacement drifts and the final ground support system installed in these drifts. The design of non-emplacement openings and their ground support systems is covered in the ''Ground Control for Non-Emplacement Drifts for LA'' (BSC 2004c)
Heat transfer analysis of the geologic disposal of spent fuel and high-level waste storage canisters

International Nuclear Information System (INIS)

Allen, G.K.

1980-08-01

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

International Nuclear Information System (INIS)

Wilson, T.; Novotny, R.

1999-01-01

The objective of this analysis is to identify issues and criteria that apply to the design of the Subsurface Emplacement Transportation System (SET). The SET consists of the track used by the waste package handling equipment, the conductors and related equipment used to supply electrical power to that equipment, and the instrumentation and controls used to monitor and operate those track and power supply systems. Major considerations of this analysis include: (1) Operational life of the SET; (2) Geometric constraints on the track layout; (3) Operating loads on the track; (4) Environmentally induced loads on the track; (5) Power supply (electrification) requirements; and (6) Instrumentation and control requirements. This analysis will provide the basis for development of the system description document (SDD) for the SET. This analysis also defines the interfaces that need to be considered in the design of the SET. These interfaces include, but are not limited to, the following: (1) Waste handling building; (2) Monitored Geologic Repository (MGR) surface site layout; (3) Waste Emplacement System (WES); (4) Waste Retrieval System (WRS); (5) Ground Control System (GCS); (6) Ex-Container System (XCS); (7) Subsurface Electrical Distribution System (SED); (8) MGR Operations Monitoring and Control System (OMC); (9) Subsurface Facility System (SFS); (10) Subsurface Fire Protection System (SFR); (11) Performance Confirmation Emplacement Drift Monitoring System (PCM); and (12) Backfill Emplacement System (BES)
Wire-rope emplacement of diagnostics systems

International Nuclear Information System (INIS)

Burden, W.L.

1982-01-01

The study reported here was initiated to determine if, with the Cable Downhole System (CDS) currently under development, there is an advantage to using continuous wire rope to lower the emplacement package to the bottom of the hole. A baseline design using two wire ropes as well as several alternatives are discussed in this report. It was concluded that the advantages of the wire-rope emplacement system do not justify the cost of converting to such a system, especially for LLNL's maximum emplacement package weights
Waste package emplacement borehole option study

International Nuclear Information System (INIS)

Streeter, W.S.

1992-03-01

This study evaluates the cost and thermal effects of various waste package emplacement configurations that differ in emplacement orientation, number of containers per borehole, and standoff distance at the potential Yucca Mountain nuclear waste repository. In this study, eight additional alternatives to the vertical and horizontal orientation options presented in the Site Characterization Plan Conceptual Design Report are considered. Typical panel layout configurations based on thermal analysis of the waste and cost estimates for design and construction, operations, and closure and decommissioning were made for each emplacement option. For the thermal analysis average waste 10 years out of reactor and the SIM code were used to determine whether the various configurations temperatures would exceed the design criteria for temperature. This study does not make a recommendation for emplacement configuration, but does provide information for comparison of alternatives
Effects of Fault Displacement on Emplacement Drifts

International Nuclear Information System (INIS)

Duan, F.

2000-01-01

The purpose of this analysis is to evaluate potential effects of fault displacement on emplacement drifts, including drip shields and waste packages emplaced in emplacement drifts. The output from this analysis not only provides data for the evaluation of long-term drift stability but also supports the Engineered Barrier System (EBS) process model report (PMR) and Disruptive Events Report currently under development. The primary scope of this analysis includes (1) examining fault displacement effects in terms of induced stresses and displacements in the rock mass surrounding an emplacement drift and (2 ) predicting fault displacement effects on the drip shield and waste package. The magnitude of the fault displacement analyzed in this analysis bounds the mean fault displacement corresponding to an annual frequency of exceedance of 10 -5 adopted for the preclosure period of the repository and also supports the postclosure performance assessment. This analysis is performed following the development plan prepared for analyzing effects of fault displacement on emplacement drifts (CRWMS M and O 2000). The analysis will begin with the identification and preparation of requirements, criteria, and inputs. A literature survey on accommodating fault displacements encountered in underground structures such as buried oil and gas pipelines will be conducted. For a given fault displacement, the least favorable scenario in term of the spatial relation of a fault to an emplacement drift is chosen, and the analysis is then performed analytically. Based on the analysis results, conclusions are made regarding the effects and consequences of fault displacement on emplacement drifts. Specifically, the analysis will discuss loads which can be induced by fault displacement on emplacement drifts, drip shield and/or waste packages during the time period of postclosure
Deep Borehole Emplacement Mode Hazard Analysis Revision 0

Energy Technology Data Exchange (ETDEWEB)

Sevougian, S. David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-08-07

This letter report outlines a methodology and provides resource information for the Deep Borehole Emplacement Mode Hazard Analysis (DBEMHA). The main purpose is identify the accident hazards and accident event sequences associated with the two emplacement mode options (wireline or drillstring), to outline a methodology for computing accident probabilities and frequencies, and to point to available databases on the nature and frequency of accidents typically associated with standard borehole drilling and nuclear handling operations. Risk mitigation and prevention measures, which have been incorporated into the two emplacement designs (see Cochran and Hardin 2015), are also discussed. A key intent of this report is to provide background information to brief subject matter experts involved in the Emplacement Mode Design Study. [Note: Revision 0 of this report is concentrated more on the wireline emplacement mode. It is expected that Revision 1 will contain further development of the preliminary fault and event trees for the drill string emplacement mode.
Spent fuel test - Climax: technical measurements. Interim report, Fiscal Year 1983

International Nuclear Information System (INIS)

Patrick, W.C.; Butkovich, T.R.; Carlson, R.C.

1984-02-01

The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted as part of the Nevada Nuclear Waste Storage Investigations. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April-May 1980. The spent-fuel canisters were retrieved and the thermal sources were de-energized in March-April 1983 when test data indicated that test objectives were met during the 3-year storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. In addition to emplacement and retrieval operations, three exchanges of spent-fuel between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and three previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the 3-1/2 year duration of the test on more than 900 channels. Data acquisition from the test is now limited to instrumentation calibration and evaluation activities. Data now available for analysis are presented here. Highlights of activities this year include a campaign of in situ stress measurements, mineralogical and petrological studies of pretest core samples, microfracture analyses of laboratory irradiated cores, improved calculations of near-field heat transfer and thermomechanical response during the final months of heating as well as during a six-month cool-down period, metallurgical analyses of selected test components, and further development of the data acquisition and data management systems. 27 references, 68 figures, 10 tables
Spent fuel test - Climax: technical measurements. Interim report, Fiscal Year 1983

Energy Technology Data Exchange (ETDEWEB)

Patrick, W.C.; Butkovich, T.R.; Carlson, R.C.; Durham, W.B.; Ganow, H.C.; Hage, G.L.; Majer, E.L.; Montan, D.N.; Nyholm, R.A.; Rector, N.L.

1984-02-01

The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted as part of the Nevada Nuclear Waste Storage Investigations. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April-May 1980. The spent-fuel canisters were retrieved and the thermal sources were de-energized in March-April 1983 when test data indicated that test objectives were met during the 3-year storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. In addition to emplacement and retrieval operations, three exchanges of spent-fuel between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and three previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the 3-1/2 year duration of the test on more than 900 channels. Data acquisition from the test is now limited to instrumentation calibration and evaluation activities. Data now available for analysis are presented here. Highlights of activities this year include a campaign of in situ stress measurements, mineralogical and petrological studies of pretest core samples, microfracture analyses of laboratory irradiated cores, improved calculations of near-field heat transfer and thermomechanical response during the final months of heating as well as during a six-month cool-down period, metallurgical analyses of selected test components, and further development of the data acquisition and data management systems. 27 references, 68 figures, 10 tables.
TVO-92 safety analysis of spent fuel disposal

International Nuclear Information System (INIS)

Vieno, T.; Hautojaervi, A.; Koskinen, L.; Nordman, H.

1993-08-01

The spent fuel from the TVO I and TVO II reactors at the Olkiluoto nuclear power plant is planned to be disposed in a repository constructed at a depth of about 500 meters in crystalline bedrock. Teollisuuden Voima Oy (TVO) has carried out preliminary site investigations for spent fuel disposal between 1987 and 1992 at five areas in Finland (Olkiluoto, Kivetty, Romuvaara, Syyry and Veitsivaara). The Safety analysis of the disposal system is presented in the report. Spent fuel will be encapsulated in composite copper-steel canisters. The canister design (ACP canister) consists of an inner container of steel as a load-bearing element and an outer container of oxygen-free copper to provide a shield against corrosion. In the repository the canisters will be emplaced in vertical holes drilled in the floors of horizontal deposition tunnels. The annulus between the canister and the rock is filled with compacted bentonite. The results of the safety analysis attest that the planned disposal system fulfils the safety requirements. Suitable places for the repository can be found at each of the five investigation sites
Thermomechanical modeling of the Spent Fuel Test-Climax

Energy Technology Data Exchange (ETDEWEB)

Butkovich, T.R.; Patrick, W.C.

1986-02-01

The Spent Fuel Test-Climax (SFT-C) was conducted to evaluate the feasibility of retrievable deep geologic storage of commercially generated spent nuclear-reactor fuel assemblies. One of the primary aspects of the test was to measure the thermomechanical response of the rock mass to the extensive heating of a large volume of rock. Instrumentation was emplaced to measure stress changes, relative motion of the rock mass, and tunnel closures during three years of heating from thermally decaying heat sources, followed by a six-month cooldown period. The calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels which were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares the results with selected measurements made during heating and cooling of the SFT-C.
Thermomechanical modeling of the Spent Fuel Test-Climax

International Nuclear Information System (INIS)

Butkovich, T.R.; Patrick, W.C.

1986-02-01

The Spent Fuel Test-Climax (SFT-C) was conducted to evaluate the feasibility of retrievable deep geologic storage of commercially generated spent nuclear-reactor fuel assemblies. One of the primary aspects of the test was to measure the thermomechanical response of the rock mass to the extensive heating of a large volume of rock. Instrumentation was emplaced to measure stress changes, relative motion of the rock mass, and tunnel closures during three years of heating from thermally decaying heat sources, followed by a six-month cooldown period. The calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels which were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares the results with selected measurements made during heating and cooling of the SFT-C
Ground Control for Non-Emplacement Drifts for LA

International Nuclear Information System (INIS)

Tang, D.

2004-01-01

The purpose of this calculation is to analyze the stability of repository non-emplacement drifts during the preclosure period, and to provide a final ground support method for non-emplacement drifts for the License Application (LA). This calculation will provide input for the development of LA documents. The scope of this calculation is limited to the non-emplacement drifts including access mains, ramps, exhaust mains, turnouts, intersections between access mains and turnouts, and intersections between exhaust mains and emplacement drifts, portals, TBM launch chambers, observation drift and test alcove in the performance confirmation (PC) facilities, etc. The calculation is limited to the non-emplacement drifts subjected to a combined loading of in-situ stress, seismic stress, and/or thermal stress. Other effects such as hydrological and chemical effects are not considered in this analysis
Intact and Degraded Component Criticality Calculations of N Reactor Spent Nuclear Fuel

International Nuclear Information System (INIS)

L. Angers

2001-01-01

The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k eff ) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR
Construction, emplacement, and retrievability (preclosure)

International Nuclear Information System (INIS)

McClain, W.

1985-01-01

Each of the three preclosure subgroups of the Construction, Emplacement, and Retrievability Working Group adopted a six-step approach to identify and assess current needs in geotechnical modeling and characterization. This approach may be summarized as follows: identify phenomena related to emplacement of high-level nuclear wastes, identify types of models which are required to calculate the phenomena, establish the input data needs for the models, assess the current availability of the models, assess the current status of documentation, verification, and validation of the models, and determine the adequacy of instrumentation and measurement techniques to (a) validate the models, where necessary, and (b) obtain input data for design. Systematic application of these six steps leads to the establishment of the research requirements for geotechnical modeling and characterization. A summary of modeling techniques which apply to the three subsequent sections on construction, emplacement, and retrievability is presented. Research needs, which apply to all preclosure activities, are summarized
Ground Control for Emplacement Drifts for SR

International Nuclear Information System (INIS)

Y. Sun

2000-01-01

This analysis demonstrates that a satisfactory ground control system can be designed for the Yucca Mountain site, and provides the technical basis for the design of ground support systems to be used in repository emplacement and non-emplacement drifts. The repository ground support design was based on analytical methods using acquired computer codes, and focused on the final support systems. A literature review of case histories, including the lessons learned from the design and construction of the ESF, the studies on the seismic damages of underground openings, and the use of rock mass classification systems in the ground support design, was conducted (Sections 6.3.4 and 6.4). This review provided some basis for determining the inputs and methodologies used in this analysis. Stability of the supported and unsupported emplacement and non-emplacement drifts was evaluated in this analysis. The excavation effects (i.e., state of the stress change due to excavation), thermal effects (i.e., due to heat output from waste packages), and seismic effects (i.e., from potential earthquake events) were evaluated, and stress controlled modes of failure were examined for two in situ stress conditions (k 0 =0.3 and 1.0) using rock properties representing rock mass categories of 1 and 5. Variation of rock mass units such as the non-lithophysal (Tptpmn) and lithophysal (Tptpll) was considered in the analysis. The focus was on the non-lithophysal unit because this unit appears to be relatively weaker and has much smaller joint spacing. Therefore, the drift stability and ground support needs were considered to be controlled by the design for this rock unit. The ground support systems for both emplacement and non-emplacement drifts were incorporated into the models to assess their performance under in situ, thermal, and seismic loading conditions. Both continuum and discontinuum modeling approaches were employed in the analyses of the rock mass behavior and in the evaluation of the

Repository emplacement costs for Al-clad high enriched uranium spent fuel

International Nuclear Information System (INIS)

McDonell, W.R.; Parks, P.B.

1994-01-01

A range of strategies for treatment and packaging of Al-clad high-enriched uranium (HEU) spent fuels to prevent or delay the onset of criticality in a geologic repository was evaluated in terms of the number of canisters produced and associated repository costs incurred. The results indicated that strategies in which neutron poisons were added to consolidated forms of the U-Al alloy fuel generally produced the lowest number of canisters and associated repository costs. Chemical processing whereby the HEU was removed from the waste form was also a low cost option. The repository costs generally increased for isotopic dilution strategies, because of the substantial depleted uranium added. Chemical dissolution strategies without HEU removal were also penalized because of the inert constituents in the final waste glass form. Avoiding repository criticality by limiting the fissile mass content of each canister incurred the highest repository costs
Instrumentation Report No. 2: identification, evaluation, and remedial actions related to transducer failures at the spent fuel test-climax

Energy Technology Data Exchange (ETDEWEB)

Patrick, W.C.; Carlson, R.C.; Rector, N.L.

1981-11-30

The Spent Fuel Test-Climax (SFT-C) is a test of the feasibility of safe and reliable short-term storage and retrieval of spent fuel from commercial nuclear reactors. In support of operational and technical goals of the test, about 850 channels of instrumentation have been installed at the SFT-C. Failure of several near-field instruments began less than six months after emplacement of 11 canisters of spent fuel and activation of six thermally similar simulators. The failed units were linear potentiometers (used to make displacement measurements) and vibrating wire stressmeters (used to make change-in-stress measurements). This report discusses the observed problems and remedial actions taken to date.
Instrumentation Report No. 2: identification, evaluation, and remedial actions related to transducer failures at the spent fuel test-climax

International Nuclear Information System (INIS)

Patrick, W.C.; Carlson, R.C.; Rector, N.L.

1981-01-01

The Spent Fuel Test-Climax (SFT-C) is a test of the feasibility of safe and reliable short-term storage and retrieval of spent fuel from commercial nuclear reactors. In support of operational and technical goals of the test, about 850 channels of instrumentation have been installed at the SFT-C. Failure of several near-field instruments began less than six months after emplacement of 11 canisters of spent fuel and activation of six thermally similar simulators. The failed units were linear potentiometers (used to make displacement measurements) and vibrating wire stressmeters (used to make change-in-stress measurements). This report discusses the observed problems and remedial actions taken to date
Comparison of risks due to HLW and SURF repositories in bedded salt

International Nuclear Information System (INIS)

Chu, M.S.Y.; Ortiz, N.R.; Wahi, K.K.

1983-01-01

A methodology was developed for use in the analysis of risks from geologic disposal of nuclear wastes. This methodology is applied to two conceptual nuclear waste repositories in bedded salt containing High-Level Waste (HLW) and Spent Un-Reprocessed Fuel (SURF), respectively. A comparison of the risk estimated from the HLW and SURF repositories is presented
The KBS concepts - General outline, present study

International Nuclear Information System (INIS)

Pusch, R.

1980-01-01

The Swedish KBS 2 concept, which concerns spent, unreprocessed reactor fuel, implies the use of an 'engineered' barrier of highly compacted Na bentonite for isolating metal canisters with the wastes from the surrounding rock. The isolating power of a barrier of this kind will be so great that it will probably be suggested for other radioactive wastes as well
Novel Emplacement Device for a Very Deep Borehole Disposal

Energy Technology Data Exchange (ETDEWEB)

Lee, Min Soo; Choi, Heui-joo; Lee, Jong Yul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

There is a worldwide attempt of HLW disposal into a very deep borehole of around 3-5 km depth with the advancement of an underground excavation technology recently. As it goes into deeper underground, the rock becomes more uniform and flawless. And then the underground water circulation system at 3-5 km depth is almost disconnected with near groundwater circulation system. The canister integrity is less important in this very deep borehole disposal system unlike a general geologic disposal system at 500 m. In the deep borehole disposal procedures, one SNF (Spent Nuclear Fuel) assembly is stored in one disposal canister (D30-40cm, H4.7-5.0m), and approximately 10-40 disposal canisters are connected axially, which parade length can leach to around 200m in maximum. The connected canister parade is lowered through a very deep borehole (D40-50cm) by emplacement devices. Therefore the connections between canisters and canister to lowering joint are very important for the safe operation of it. The well-known connection method between canisters is Threaded Coupled Connection method, in which releasing of the connection is almost impossible after thread fastening in the borehole. The novel joint device suggested in this paper can accommodate a canister emplacement and retrieval in the borehole disposal process. The joint can be lowered by bound to a drilling pipe, or high tension cable along 3-5 km distance. This novel device can cope with an accidental event easily without any joint head change. When canisters are damaged or stuck on the borehole wall during their descending, the canisters in trouble can be retrieved simply by the control of a lifting speed.
Thermal analysis for a spent reactor fuel storage test in granite

International Nuclear Information System (INIS)

Montan, D.N.

1980-09-01

A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy's Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m 3 /s and an ambient rock temperature of 23 0 C, the maximum calculated rock temperature (near the center of the heat source) is about 100 0 C while the maximum air temperature in the drift is around 40 0 C. This ventilation (1 m 3 /s through the main drift and 1/2 m 3 /s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage
Concept of Operations for Waste Transport, Emplacement, and Retrieval

International Nuclear Information System (INIS)

Raczka, Norman T.

2001-01-01

The preparation of this technical report has two objectives. The first objective is to discuss the base case concepts of waste transport, emplacement, and retrieval operations and evaluate these operations relative to a lower-temperature repository design. Aspects of the operations involved in waste transport, emplacement and retrieval may be affected by the lower-temperature operating schemes. This report evaluates the effects the lower-temperature alternatives may have on the operational concepts involved in emplacing and retrieving waste. The second objective is to provide backup material for the design description, in a traceable and defensible format, for Section 2 of the Waste Emplacement/Retrieval System Description Document
Review of potential subsurface permeable barrier emplacement and monitoring technologies

International Nuclear Information System (INIS)

Riggsbee, W.H.; Treat, R.L.; Stansfield, H.J.; Schwarz, R.M.; Cantrell, K.J.; Phillips, S.J.

1994-02-01

This report focuses on subsurface permeable barrier technologies potentially applicable to existing waste disposal sites. This report describes candidate subsurface permeable barriers, methods for emplacing these barriers, and methods used to monitor the barrier performance. Two types of subsurface barrier systems are described: those that apply to contamination.in the unsaturated zone, and those that apply to groundwater and to mobile contamination near the groundwater table. These barriers may be emplaced either horizontally or vertically depending on waste and site characteristics. Materials for creating permeable subsurface barriers are emplaced using one of three basic methods: injection, in situ mechanical mixing, or excavation-insertion. Injection is the emplacement of dissolved reagents or colloidal suspensions into the soil at elevated pressures. In situ mechanical mixing is the physical blending of the soil and the barrier material underground. Excavation-insertion is the removal of a soil volume and adding barrier materials to the space created. Major vertical barrier emplacement technologies include trenching-backfilling; slurry trenching; and vertical drilling and injection, including boring (earth augering), cable tool drilling, rotary drilling, sonic drilling, jetting methods, injection-mixing in drilled holes, and deep soil mixing. Major horizontal barrier emplacement technologies include horizontal drilling, microtunneling, compaction boring, horizontal emplacement, longwall mining, hydraulic fracturing, and jetting methods
Thermal simulation of drift emplacement. Geotechnical and geophysical investigations in and around backfilled galleries

International Nuclear Information System (INIS)

Schneefub, J.U.; Gommlich, G.E.

1988-01-01

The concept for the direct disposal of spent fuel in rock salt foresees the emplacement of large waste canisters on the floor of a disposal gallery. Subsequent to emplacement the drift is backfilled with salt grit. In a demonstration test six cylindrical containments are to be emplaced within distances of 3 m from each other in two parallel galleries of 14 m 2 , separated by a pillar of 10 m thickness. They will be heated up by a power output of approx. 1 kW/m to 200 degree C surface temperature by electrical heaters. The thermal and mechanical response of the salt rock and the backfilling to the artificial heating is to be investigated as follows: (1) measurement of the temperature field at the contained surface, in the backfill and in the rock salt; (2) deformation measurements of the salt rock around the heated drifts; (3) measurements of tunnel convergence in heated and unheated sections; (4) compaction measurements of the backfilling in heated and unheated areas; (5) measurement of rock stresses in areas very close to the galleries; and (6) pressure measurements in the backfilling, between backfilling and rock, and between backfilling and containers. The rock burst monitoring system of the Asse salt mine will be expanded to this special field of the mine to detect seismic events due to thermomechanical effects. Another seismoacoustic system will be installed to observe the compaction of the backfilling. This system must be calibrated for the relation between the density and velocities of seismic waves. It is planned to monitor the density by gamma-gamma log measurements. The changes in overall density during the entire experiment will be observed by gravimeter measurements of high precision
Implementation of the full-scale emplacement (FE) experiment at the Mont Terri rock laboratory

Energy Technology Data Exchange (ETDEWEB)

Müller, H.R.; Garitte, B.; Vogt, T.; and others

2017-04-15

Opalinus Clay is currently being assessed as the host rock for a deep geological repository for high-level and low- and intermediate-level radioactive wastes in Switzerland. Within this framework, the 'Full-Scale Emplacement' (FE) experiment was initiated at the Mont Terri rock laboratory close to the small town of St-Ursanne in Switzerland. The FE experiment simulates, as realistically as possible, the construction, waste emplacement, backfilling and early post-closure evolution of a spent fuel/vitrified high-level waste disposal tunnel according to the Swiss repository concept. The main aim of this multiple heater test is the investigation of repository-induced thermo-hydro-mechanical (THM) coupled effects on the host rock at this scale and the validation of existing coupled THM models. For this, several hundred sensors were installed in the rock, the tunnel lining, the bentonite buffer, the heaters and the plug. This paper is structured according to the implementation timeline of the FE experiment. It documents relevant details about the instrumentation, the tunnel construction, the production of the bentonite blocks and the highly compacted 'granulated bentonite mixture' (GBM), the development and construction of the prototype 'backfilling machine' (BFM) and its testing for horizontal GBM emplacement. Finally, the plug construction and the start of all 3 heaters (with a thermal output of 1350 Watt each) in February 2015 are briefly described. In this paper, measurement results representative of the different experimental steps are also presented. Tunnel construction aspects are discussed on the basis of tunnel wall displacements, permeability testing and relative humidity measurements around the tunnel. GBM densities achieved with the BFM in the different off-site mock-up tests and, finally, in the FE tunnel are presented. Finally, in situ thermal conductivity and temperature measurements recorded during the first heating months
Spent fuel test-climax: a test of geologic storage of high-level waste in granite

International Nuclear Information System (INIS)

Ramspott, L.D.; Ballou, L.B.; Patrick, W.C.

1981-01-01

A test of retrievable geologic storage of spent fuel assemblies from an operating commercial nuclear reactor is underway at the Nevada Test Site (NTS) of the US Department of Energy. This generic test is located 420 m below the surface in the Climax granitic stock. Eleven canisters of spent fuel approximately 2.5 years out of reactor core (about 1.6 kW/canister thermal output) were emplaced in a storage drift along with 6 electrical simulator canisters. Two adjacent drifts contain electrical heaters, which are operated to simulate within the test array the thermal field of a large repository. Fuel was loaded during April to May 1980 and initial results of the test will be presented
The full-scale Emplacement (FE) Experiment at the Mont Terri URL

International Nuclear Information System (INIS)

Mueller, H.R.; Weber, H.P.; Koehler, S.; Vogt, T.; Vietor, T.

2012-01-01

Document available in extended abstract form only. The Full-Scale Emplacement (FE) Experiment at the Mont Terri underground research laboratory (URL) is a full-scale heater test in a clay-rich formation. It simulates the construction, waste emplacement and backfilling of a spent fuel (SF) / vitrified high-level waste (HLW) repository tunnel as realistically as possible. The entire experiment implementation as well as the post-closure THM(C) evolution will be monitored using several hundred sensors. These are distributed in the host rock in the near- and far-field, the tunnel lining, the engineered barrier system and on the heaters. The aim of this experiment is to investigate HLW repository-induced thermo-hydro-mechanical (THM) coupled effects on the host rock and the validation of existing coupled THM models. A further aim is the verification of the technical feasibility of constructing a 50 m repository section at full scale with all relevant components using standard industrial equipment. Finally, the experiment will demonstrate the canister and buffer emplacement procedures for underground conditions based on the Swiss disposal concept. Experimental layout The FE experiment is based on the Swiss disposal concept for SF / HLW. The 50 m long test gallery, at the end of the former MB test tunnel in the Mont Terri URL, will be realised with a diameter of approx. 3 m. In the experiment gallery, 3 heaters with dimensions similar to those of waste canisters will be emplaced on top of abutments built of bentonite blocks. The remaining space will be backfilled with compacted bentonite pellets. The experiment will be sealed off towards the start niche with a concrete plug holding the buffer in place and reducing air and water fluxes. The first scoping calculations and design modelling for the 'far-field' instrumentation have been completed; these works have been carried out using CodeBRIGHT and the multiphase flow simulator TOUGH2. With an initial heat output of 1500 W
Room at the Mountain: Estimated Maximum Amounts of Commercial Spent Nuclear Fuel Capable of Disposal in a Yucca Mountain Repository

International Nuclear Information System (INIS)

Kessler, John H.; Kemeny, John; King, Fraser; Ross, Alan M.; Ross, Benjamen

2006-01-01

The purpose of this paper is to present an initial analysis of the maximum amount of commercial spent nuclear fuel (CSNF) that could be emplaced into a geological repository at Yucca Mountain. This analysis identifies and uses programmatic, material, and geological constraints and factors that affect this estimation of maximum amount of CSNF for disposal. The conclusion of this initial analysis is that the current legislative limit on Yucca Mountain disposal capacity, 63,000 MTHM of CSNF, is a small fraction of the available physical capacity of the Yucca Mountain system assuming the current high-temperature operating mode (HTOM) design. EPRI is confident that at least four times the legislative limit for CSNF (∼260,000 MTHM) can be emplaced in the Yucca Mountain system. It is possible that with additional site characterization, upwards of nine times the legislative limit (∼570,000 MTHM) could be emplaced. (authors)
Longevity of Emplacement Drift Ground Support Materials, Rev. 01

International Nuclear Information System (INIS)

David H. Tang

2000-01-01

The purpose of this analysis is to evaluate the factors affecting the longevity of emplacement drift ground support materials and to develop a basis for the selection of materials for ground support that will function throughout the preclosure period of a potential repository at Yucca Mountain. The Development Plan (DP) for this analysis is given in Longevity of Emplacement Drift Ground Support Materials (CRWMS M and O 1999a). The objective of this analysis is to update the previous analysis (CRWMS M and O 2000a) to account for related changes in the Ground Control System Description Document (CRWMS M and O 2000b), the Monitored Geologic Repository Project Description Document (CRWMS M and O 1999b), and in environmental conditions, and to provide updated information on candidate ground support materials. Candidate materials for ground support are carbon steel and cement grout. Steel is mainly used for steel sets, lagging, channel, rock bolts, and wire mesh. Cement grout is only considered in the case of grouted rock bolts. Candidate materials for the emplacement drift invert are carbon steel and crushed rock ballast. Materials are evaluated for the repository emplacement drift environment based on the updated thermal loading condition and waste package design. The analysis consists of the following tasks: (1) Identify factors affecting the longevity of ground support materials for use in emplacement drifts; (2) Review existing documents concerning the behavior of candidate ground support materials during the preclosure period; (3) Evaluate impacts of temperature and radiation effects on mechanical and thermal properties of steel. Assess corrosion potential of steel at emplacement drift environment; (4) Evaluate factors affecting longevity of cement grouts for fully grouted rock bolt system. Provide updated information on cement grout mix design for fully grouted rock bolt system; and (5) Evaluate longevity of materials for the emplacement drift invert
Longevity of Emplacement Drift Ground Support Materials

International Nuclear Information System (INIS)

D.H.Tang

2001-01-01

The purpose of this analysis is to evaluate the factors affecting the longevity of emplacement drift ground support materials and to develop a basis for the selection of materials for ground support that will function throughout the preclosure period of a potential repository at Yucca Mountain. REV 01 ICN 01 of this analysis is developed in accordance with AP-3.10Q, Analyses and Models, Revision 2, ICN 4, and prepared in accordance with the Technical Work Plan for Subsurface Design Section FY 01 Work Activities (CRWMS M and O 2001a). The objective of this analysis is to update the previous analysis (CRWMS M and O 2000a) to account for related changes in the Ground Control System Description Document (CRWMS M and O 2000b), the Monitored Geologic Repository Project Description Document, which is included in the Requirements and Criteria for Implementing a Repository Design that can be Operated Over a Range of Thermal Modes (BSC 2001), input information, and in environmental conditions, and to provide updated information on candidate ground support materials. Candidate materials for ground support are carbon steel and cement grout. Steel is mainly used for steel sets, lagging, channel, rock bolts, and wire mesh. Cement grout is only considered in the case of grouted rock bolts. Candidate materials for the emplacement drift invert are carbon steel and granular natural material. Materials are evaluated for the repository emplacement drift environment based on the updated thermal loading condition and waste package design. The analysis consists of the following tasks: (1) Identify factors affecting the longevity of ground support materials for use in emplacement drifts. (2) Review existing documents concerning the behavior of candidate ground support materials during the preclosure period. (3) Evaluate impacts of temperature and radiation effects on mechanical and thermal properties of steel. Assess corrosion potential of steel at emplacement drift environment. (4
WIPP waste package testing on simulated DHLW: emplacement

International Nuclear Information System (INIS)

Molecke, M.A.

1984-01-01

Several series of simulated (nonradioactive) defense high-level waste (DHLW) package tests have been emplaced in the WIPP, a research and development facility authorized to demonstrate the safe disposal of defense-related wastes. The primary purpose of these 3-to-7 year duration tests is to evaluate the in situ materials performance of waste package barriers (canisters, overpacks, backfills, and nonradioactive DHLW glass waste form) for possible future application to a licensed waste repository in salt. This paper describes all test materials, instrumentation, and emplacement and testing techniques, and discusses progress of the various tests. These tests are intended to provide information on materials behavior (i.e., corrosion, metallurgical and geochemical alterations, waste form durability, surface interactions, etc.), as well as comparison between several waste package designs, fabrications details, and actual costs. These experiments involve 18 full-size simulated DHLW packages (approximately 3.0 m x 0.6 m diameter) emplaced in vertical boreholes in the salt drift floor. Six of the test packages contain internal electrical heaters (470 W/canister), and were emplace under approximately reference DHLW repository conditions. Twelve other simulated DHLW packages were emplaced under accelerated-aging or overtest conditions, including the artificial introduction of brine, and a thermal loading approximately three to four times higher than reference. Eight of these 12 test packages contain 1500 W/canister electrical heaters; the other four are filled with DHLW glass. 9 refs., 1 fig
Systems impacts of spent fuel disassembly alternatives

International Nuclear Information System (INIS)

1984-07-01

Three studies were completed to evaluate four alternatives to the disposal of intact spent fuel assemblies in a geologic repository. A preferred spent fuel waste form for disposal was recommended on consideration of (1) package design and fuel/package interaction, (2) long-term, in-repository performance of the waste form, and (3) overall process performance and costs for packaging, handling, and emplacement. The four basic alternative waste forms considered were (1) end fitting removal, (2) fission gas venting, (3) disassembly and close packing, and (4) shearing/immobilization. None of the findings ruled out any alternative on the basis of waste package considerations or long-term performance of the waste form. The third alternative offers flexibility in loading that may prove attractive in the various geologic media under consideration, greatly reduces the number of packages, and has the lowest unit cost. These studies were completed in October, 1981. Since then Westinghouse Electric Corporation and the Office of Nuclear Waste Isolation have completed studies in related fields. This report is now being published to provide publicly the background material that is contained within. 47 references, 28 figures, 31 tables
Design management and stress analysis of a circular rock tunnel and emplacement holes for storage of spent nuclear fuel

International Nuclear Information System (INIS)

Kandalaft-Ladkany, N.; Wyman, R.V.

1992-01-01

This paper discusses a critical path method (CPM) diagram and logic net which are used for the design cycle of the rock tunnel system for a high level nuclear waste repository. In the analysis the design tunnel is subjected to pre-existing temperature and overburden loads at time of construction. high thermal stresses develop later due to the long term influx of heat from the canisters stored in vertical emplacement holes. Results indicate that thermal stresses reach a critical level for the rock in the vicinity of the canisters which could lead to local collapse of the rock and damage to the canisters
Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

International Nuclear Information System (INIS)

Hardin, Ernest; Matteo, Edward N.; Hadgu, Teklu

2016-01-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all "enclosed,"whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative "open"modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if "enclosed"concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

Energy Technology Data Exchange (ETDEWEB)

Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Matteo, Edward N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-01-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).
Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design.

Energy Technology Data Exchange (ETDEWEB)

Cochran, John R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-07-01

This report presents conceptual design information for a system to handle and emplace packages containing radioactive waste, in boreholes 16,400 ft deep or possibly deeper. Its intended use is for a design selection study that compares the costs and risks associated with two emplacement methods: drill-string and wireline emplacement. The deep borehole disposal (DBD) concept calls for siting a borehole (or array of boreholes) that penetrate crystalline basement rock to a depth below surface of about 16,400 ft (5 km). Waste packages would be emplaced in the lower 6,560 ft (2 km) of the borehole, with sealing of appropriate portions of the upper 9,840 ft (3 km). A deep borehole field test (DBFT) is planned to test and refine the DBD concept. The DBFT is a scientific and engineering experiment, conducted at full-scale, in-situ, without radioactive waste. Waste handling operations are conceptualized to begin with the onsite receipt of a purpose-built Type B shipping cask, that contains a waste package. Emplacement operations begin when the cask is upended over the borehole, locked to a receiving flange or collar. The scope of emplacement includes activities to lower waste packages to total depth, and to retrieve them back to the surface when necessary for any reason. This report describes three concepts for the handling and emplacement of the waste packages: 1) a concept proposed by Woodward-Clyde Consultants in 1983; 2) an updated version of the 1983 concept developed for the DBFT; and 3) a new concept in which individual waste packages would be lowered to depth using a wireline. The systems described here could be adapted to different waste forms, but for design of waste packaging, handling, and emplacement systems the reference waste forms are DOE-owned high- level waste including Cs/Sr capsules and bulk granular HLW from fuel processing. Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design July 23, 2015 iv ACKNOWLEDGEMENTS This report has
Emplacement and stemming of nuclear explosives for Plowshare applications

Energy Technology Data Exchange (ETDEWEB)

Cramer, J L [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

1970-05-15

This paper will discuss the various methods used for emplacement and design considerations that must be taken into account when the emplacement and stemming method is selected. The step-by-step field procedure will not be discussed in this paper. The task of emplacing and stemming the nuclear explosive is common to all Plowshare experiments today. All present-day applications of a nuclear explosive for Plowshare experiments require that the detonation take place some distance below the surface of the ground. This is normally done by lowering the explosive into an emplacement hole to a desired depth and then backfilling the hole with a suitable stemming material. At first glance it scenes like a very straightforward, simple task to perform. It would appear to be a task that could become a standard procedure for all experiments; however, this is not the case. In actuality, the emplacement and stemming of a nuclear explosive must almost be a custom design. It varies with the application of the experiment, i.e., cratering or underground engineering. It also varies with the condition of the hole, the available equipment to do the job, the actual purpose of the stemming, possible postshot reentry, hydrology, geology, and future production. A very important item that must always be considered is the protection of the firing and signal cables during the downhole and stemming operation. Each of these things must be considered; ignoring any one of them could jeopardize one of the objectives of the experiment or perhaps even the experiment itself. It should be emphasized that for a multiple-shot program such as would be used to develop a gas field where the geology, depths of burial etc. are the same, the emplacement and stemming operation would be standardized, as would all other parts of the program. However, for individual experiments in totally different areas, complete standardization of the emplacement and stemming is impossible.
Emplacement and stemming of nuclear explosives for Plowshare applications

International Nuclear Information System (INIS)

Cramer, J.L.

1970-01-01

This paper will discuss the various methods used for emplacement and design considerations that must be taken into account when the emplacement and stemming method is selected. The step-by-step field procedure will not be discussed in this paper. The task of emplacing and stemming the nuclear explosive is common to all Plowshare experiments today. All present-day applications of a nuclear explosive for Plowshare experiments require that the detonation take place some distance below the surface of the ground. This is normally done by lowering the explosive into an emplacement hole to a desired depth and then backfilling the hole with a suitable stemming material. At first glance it scenes like a very straightforward, simple task to perform. It would appear to be a task that could become a standard procedure for all experiments; however, this is not the case. In actuality, the emplacement and stemming of a nuclear explosive must almost be a custom design. It varies with the application of the experiment, i.e., cratering or underground engineering. It also varies with the condition of the hole, the available equipment to do the job, the actual purpose of the stemming, possible postshot reentry, hydrology, geology, and future production. A very important item that must always be considered is the protection of the firing and signal cables during the downhole and stemming operation. Each of these things must be considered; ignoring any one of them could jeopardize one of the objectives of the experiment or perhaps even the experiment itself. It should be emphasized that for a multiple-shot program such as would be used to develop a gas field where the geology, depths of burial etc. are the same, the emplacement and stemming operation would be standardized, as would all other parts of the program. However, for individual experiments in totally different areas, complete standardization of the emplacement and stemming is impossible
Mechanisms Of Saucer-Shaped Sill Emplacement: Insight From Experimental Modeling

Science.gov (United States)

Galland, O.; Planke, S.; Malthe-Sørenssen, A.; Polteau, S.; Svensen, H.; Podladchikov, Y. Y.

2006-12-01

It has been recently demonstrated that magma intrusions in sedimentary basins had a strong impact on petroleum systems. Most of these intrusions are sills, and especially saucer-shaped sills. These features can be observed in many sedimentary basins (i.e. the Karoo basin, South Africa; the Norwegian and North Sea; the Tunguska basin, Siberia; the Neuquén basin in Argentina). The occurrence of such features in so various settings suggests that their emplacement results from fundamental processes. However, the mechanisms that govern their formation remain poorly constrained. Experiments were conducted to simulate the emplacement of saucer-shaped magma intrusions in sedimentary basins. The model rock and magma were fine-grained silica flour and molten vegetable oil, respectively. This modeling technique allows simultaneous simulation of magma emplacement and brittle deformation at a basin scale. For our purpose, we performed our experiments without external deformation. During the experiments, the oil was injected horizontally at constant flow rate within the silica flour. Then the oil initially emplaced in a sill, whereas the surface of the model inflated into a smooth dome. Subsequently, the oil propagated upwards along inclined sheets, finally reaching the surface at the edge of the dome. The resulting geometries of the intrusions were saucer-shaped sills. Then the oil solidified, and the model was cut in serial cross-sections through which the structures of the intrusive body and of the overburden can be observed. In order to constraint the processes governing the emplacement of such features, we performed a parametric study based on a set of experiments in which we systematically varied parameters such as the depth of emplacement and the injection flow rate of the oil. Our results showed that saucer diameters are larger at deeper level of emplacement. Opposite trend was obtained with varying injection flow rates. Based on our results, we conducted a detailed
National waste terminal storage conceptual reference repository description

International Nuclear Information System (INIS)

Odgers, I.L.; Collings, J.L.

1981-01-01

The conceptual reference repository description (CRRD) discusses, from a conceptual engineering standpoint, the structures, systems, equipment, and operations necessary to: (1) receive unreprocessed spent fuel assemblies in standard casks; (2) unload these assemblies; (3) place them in canisters; (4) transport the canisters to underground storage locations in the salt dome; and (5) place the canisters in terminal storage. The CRRD also elaborates on the concepts for retrieval and recovery of the spent fuel after burial and describes the development of the shafts and the underground areas, as well as the supporting operational utility and administrative features of the repository
Dry well storage of spent LWBR fuel

International Nuclear Information System (INIS)

Christensen, A.B.; Fielding, K.D.

1985-01-01

Recently, 50 dry wells were constructed at the Idaho Chemical Processing Plant (ICPP) to temporarily store the Light Water Breeder Reactor (LWBR) fuel. Over 400 dry wells of the same design are projected to be constructed in the next 5 yr at the ICPP to store unreprocessible fuels until a permanent repository becomes available. This summary describes the LWBR fuel storage dry wells and the enhancements made over the Peach Bottom fuel and Fermi blanket dry wells that have been in use for up to 4 yr. Dry well storage at the ICPP has historically been found to be a safe and efficient method of temporary fuel storage. The LWBR dry wells should be more reliable than the original dry wells and provide data not previously available
Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

Energy Technology Data Exchange (ETDEWEB)

Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Matteo, Edward N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-12-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).
Estimates of relative areas for the disposal in bedded salt of LWR wastes from alternative fuel cycles

International Nuclear Information System (INIS)

Lincoln, R.C.; Larson, D.W.; Sisson, C.E.

1978-01-01

The relative mine-level areas (land use requirements) which would be required for the disposal of light-water reactor (LWR) radioactive wastes in a hypothetical bedded-salt formation have been estimated. Five waste types from alternative fuel cycles have been considered. The relative thermal response of each of five different site conditions to each waste type has been determined. The fuel cycles considered are the once-through (no recycle), the uranium-only recycle, and the uranium and plutonium recycle. The waste types which were considered include (1) unreprocessed spent reactor fuel, (2) solidified waste derived from reprocessing uranium oxide fuel, (3) plutonium recovered from reprocessing spent reactor fuel and doped with 1.5% of the accompanying waste from reprocessing uranium oxide fuel, (4) waste derived from reprocessing mixed uranium/plutonium oxide fuel in the third recycle, and (5) unreprocessed spent fuel after three recycles of mixed uranium/plutonium oxide fuels. The relative waste-disposal areas were determined from a calculated value of maximum thermal energy (MTE) content of the geologic formations. Results are presented for each geologic site condition in terms of area ratios. Disposal area requirements for each waste type are expressed as ratios relative to the smallest area requirement (for waste type No. 2 above). For the reference geologic site condition, the estimated mine-level disposal area ratios are 4.9 for waste type No. 1, 4.3 for No. 3, 2.6 for No. 4, and 11 for No. 5
Alternatives for water basin spent fuel storage: executive summary and comparative evaluation

International Nuclear Information System (INIS)

Viebrock, J.M.

1979-09-01

A five part report identifies and evaluates alternatives to conventional methods for water basin storage of irradiated light water reactor fuel assemblies (spent fuel). A recommendation is made for development or further evaluation of one attractive alternative: Proceed to develop fuel disassembly with subsequent high density storage of fuel pins (pin storage). The storage alternatives were evaluated for emplacement at reactor, in existing away-from-reactor storage facilities and in new away-from-reactor facilities. In the course of the study, the work effort necessarily extended beyond the pool wall in scope to properly assess the affects of storage alternatives on AFT systems
Automated emplacement and retrieval of hazardous waste

International Nuclear Information System (INIS)

Slocum, A.H.; Hou, W.M.

1987-01-01

The design of several dedicated machines to perform simple tasks often results in higher system reliability and efficiency than the design of a single, multifunctional machine. Similarly, a reliable system for emplacement and retrieval of nuclear waste can be realized if emplacement/retrieval operations are decomposed into a well-defined series of independent tasks. The basic methodology is to design a system that eliminates contact between the waste package and the vehicle in the event of machine failure. The disabled vehicle can then be withdrawn to a safe location, repaired, and set back to resume normal operation
Phenomenological description reference concept (Spent Fuel - Copper - Iron - Bentonite - Granite)

International Nuclear Information System (INIS)

Lehikoinen, J.; Nordman, H.

2008-12-01

This report gives a broad overview of the salient processes and associated time frames and uncertainties pertaining to the long-term normal evolution of the near-field of a deep repository of the KBS-3V type. By normal evolution, reference is made to the description of the evolution of a disposal system subject to the assumption that all the spent fuel canisters remain intact following emplacement and repository closure for at least the design corrosion lifetime of the canister overpack in the expected repository conditions. In this connection, the effects of future glaciations, canister manufacturing defects and small-scale tectonic activity are briefly touched upon. The roles of the different repository near-field barriers in providing safety are briefly noted. The description of the normal evolution of a KBS-3V-type repository near-field conditions external to the unbreached spent fuel canister has been realized in terms of processes rather than timescales. Separate chapters have been devoted to the thermal, hydraulic, mechanical and chemical/biological evolution of the repository system, with due attention to some of the couplings between them. (orig.)
Rooted in two Countries? Migrant Heritage, emplacement and the ‘Canon van Nederland’.

NARCIS (Netherlands)

van Faassen, M.; Hoekstra, F.G.

2017-01-01

Rooted in two Countries? Migrant Heritage, emplacement and the ‘Canon van Nederland’. In the cultural heritage sector theorizing emplacement is considered to be vital for identity: “people need to anchor there identity concretely to a location (emplacement)” (Grever&Van Boxtel, 2014). The question
Melt-Dilute Spent Nuclear Fuel Form Criticality Summary Report; FINAL

International Nuclear Information System (INIS)

Vinson, D.W.

2002-01-01

Criticality analysis of the proposed Melt-Dilute (MD) form of aluminum-based spent nuclear fuel (SNF), under geologic repository conditions, was performed following the methodology, documented in the Disposal Criticality Analysis Methodology Topical Report. This methodology evaluates the potential for nuclear criticality as determined by the composition of the waste and its geometry, namely waste form configuration, including presence of moderator, reflecting structural material, and neutron absorbers. The initial emplaced configuration of the SNF form is a dry package placed in a mined repository passageway. Criticality calculations show that even with waste package failure, followed by degradation of material within the waste package and potential loss of neutron absorber materials, sub-critical conditions can be maintained
Evaluation of alternative spent fuel waste package concepts for a repository in Basalt

International Nuclear Information System (INIS)

Hall, G.V.B.; Nair, B.R.

1986-01-01

The United States government has established a program for the disposal of spent nuclear fuel and high-level radioactive waste. The Nuclear Waste Policy Act (NWPA) of 1982 requires the first nuclear waste repository to begin receiving high-level radioactive waste in 1998. One of the potentially acceptable sites currently being evaluated is the Hanford Site in the Pasco Basin in the state of Washington where the host rock is basalt. Under the direction of the United States Department of Energy (DOE), Rockwell International's Rockwell Hanford Operations (RHO) has initiated the Basalt Waste Isolation Project (BWIP). The BWIP must design waste packages for emplacement in the repository. As part of the BWIP waste package development program, several alternative designs were considered for the disposal of spent nuclear fuel. This paper describes the concepts that were evaluated, the criteria that was developed for judging their relative merits, and the methodology that was employed. The results of the evaluation show that a Pipe-In-Tunnel design, which uses a long carbon steel pipe for the containment barrier for multiple packages of consolidated spent fuel, has the highest rating. Other designs which had high ratings are also discussed
Investigations of possibilities to dispose of spent nuclear fuel in Lithuania: a model case. Volume 2, Concept of Repository in Crystalline Rocks

International Nuclear Information System (INIS)

Motiejunas, S.; Poskas, P.

2005-01-01

The aim is to present the generic repository concept in crystalline rocks in Lithuania and cost assessment of the disposal of spent nuclear fuel and long-lived intermediate level waste in this repository. Due to limited budget of this project the repository concept development for Lithuania was based mostly on the experience of foreign countries. In this Volume a review of the existing information on disposal concept in crystalline rocks from various countries is presented. Described repository concept for crystalline rocks in Lithuania covers repository layout, backfill, canister, construction materials and auxiliary buildings. Costs calculations for disposal of spent nuclear fuel and long-lived intermediate-level wastes from Ignalina NPP are presented too. Thermal, criticality and other important disposal evaluations for RBMK-1500 spent nuclear fuel emplaced in copper canister were performed and described
Handling encapsulated spent fuel in a geologic repository environment

International Nuclear Information System (INIS)

Ballou, L.B.

1983-02-01

In support of the Spent Fuel Test-Climate at the U.S. Department of Energy's Nevada Test Site, a spent-fuel canister handling system has been designed, deployed, and operated successfully during the past five years. This system transports encapsulated commercial spent-fuel assemblies between the packaging facility and the test site (approx. 100 km), transfers the canisters 420 m vertically to and from a geologic storage drift, and emplaces or retrieves the canisters from the storage holes in the floor of the drift. The spent-fuel canisters are maintained in a fully shielded configuration at all times during the handling cycle, permitting manned access at any time for response to any abnormal conditions. All normal operations are conducted by remote control, thus assuring as low as reasonably achievable exposures to operators; specifically, we have had no measurable exposure during 30 canister transfer operations. While not intended to be prototypical of repository handling operations, the system embodies a number of concepts, now demonstrated to be safe, reliable, and economical, which may be very useful in evaluating full-scale repository handling alternatives in the future. Among the potentially significant concepts are: Use of an integral shielding plug to minimize radiation streaming at all transfer interfaces. Hydraulically actuated transfer cask jacking and rotation features to reduce excavation headroom requirements. Use of a dedicated small diameter (0.5 m) drilled shaft for transfer between the surface and repository workings. A wire-line hoisting system with positive emergency braking device which travels with the load. Remotely activated grapples - three used in the system - which are insensitive to load orientation. Rail-mounted underground transfer vehicle operated with no personnel underground
Cryovolcanic Emplacement of Domes on Europa

Science.gov (United States)

Quick, Lynnae C.; Glaze, Lori S.; Baloga, Stephen M.

2016-01-01

Here we explore the hypothesis that certain domes on Europa may have been produced by the extrusion of viscous cryolavas. A new mathematical method for the emplacement and relaxation of viscous lava domes is presented and applied to putative cryovolcanic domes on Europa. A similarity solution approach is applied to the governing equation for fluid flow in a cylindrical geometry, and dome relaxation is explored assuming a volume of cryolava has been rapidly emplaced onto the surface. Nonphysical sin- gularities inherent in previous models for dome relaxation have been eliminated, and cryolava cooling is represented by a time-variable viscosity. We find that at the onset of relaxation, bulk kinematic viscosities may lie in the range between 10(exp 3) and 10(exp 6) sq m/s, while the actual fluid lava viscosity may be much lower. Plausible relaxation times to form the domes, which are linked to bulk cryolava rheology, are found to range from 3.6 days to 7.5 years. We find that cooling of the cryolava, while dominated by conduction through an icy skin, should not prevent fluids from advancing and relaxing to form domes within the timescales considered. Determining the range of emplacement conditions for putative cryolava domes will shed light on Europa's resurfacing history. In addition, the rheologies and compositions of erupted cryolavas have implications for subsurface cryomagma ascent and local surface stress conditions on Europa.
Spent fuel test - Climax: technical measurements. Interim report, fiscal year 1981

International Nuclear Information System (INIS)

Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

1982-01-01

The Spent Fuel Test-Climax (SFT-C) is located 420 m below surface in the Climax granite stock on the Nevada Test Site. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized from April to May 1980, initiating the 3- to 5-year-duration test. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Technical objectives of the test led to development of a technical measurements program, which is the subject of this report. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 1-1/2 years of the test on more than 900 channels. Much of the acquired data are now available for analysis and are presented here. Highlights of activities this year include completion of site characterization field work, major modifications to the data acquisition and the management systems, and the addition of instrument evaluation as an explicit objective of the test
Spent fuel test - Climax: technical measurements. Interim report, fiscal year 1981

Energy Technology Data Exchange (ETDEWEB)

Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

1982-04-30

The Spent Fuel Test-Climax (SFT-C) is located 420 m below surface in the Climax granite stock on the Nevada Test Site. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized from April to May 1980, initiating the 3- to 5-year-duration test. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Technical objectives of the test led to development of a technical measurements program, which is the subject of this report. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 1-1/2 years of the test on more than 900 channels. Much of the acquired data are now available for analysis and are presented here. Highlights of activities this year include completion of site characterization field work, major modifications to the data acquisition and the management systems, and the addition of instrument evaluation as an explicit objective of the test.

Development of an international safeguards approach to the final disposal of spent fuel in geological repositories

International Nuclear Information System (INIS)

Murphey, W.M.; Moran, B.W.; Fattah, A.

1996-01-01

The International Atomic Energy Agency (IAEA) is currently pursuing development of an international safeguards approach for the final disposal of spent fuel in geological repositories through consultants meetings and through the Program for Development of Safeguards for Final Disposal of Spent Fuel in Geological Repositories (SAGOR). The consultants meetings provide policy guidance to IAEA; SAGOR recommends effective approaches that can be efficiently implemented by IAEA. The SAGOR program, which is a collaboration of eight Member State Support Programs (MSSPs), was initiated in July 1994 and has identified 15 activities in each of three areas (i.e. conditioning facilities, active repositories, and closed repositories) that must be performed to ensure an efficient, yet effective safeguards approach. Two consultants meetings have been held: the first in May 1991 and the last in November 1995. For nuclear materials emplaced in a geological repository, the safeguards objectives were defined to be (1) to detect the diversion of spent fuel, whether concealed or unconcealed, from the repository and (2) to detect undeclared activities of safeguards concern (e.g., tunneling, underground reprocessing, or substitution in containers)
Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

Directory of Open Access Journals (Sweden)

Herrero J.J.

2017-01-01

Full Text Available In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.
EMPLACEMENT GANTRY ITS STANDARDS IDENTIFICATION STUDY

International Nuclear Information System (INIS)

Voegele, M.

2005-01-01

To date, the project has established ITS performance requirements for SSCs based on identification and categorization of event sequences that may result in a radiological release. These performance requirements are defined within the NSDB. Further, SSCs credited with performing safe functions are classified as ITS. In turn, perform confirmation for these SSCs is sought through the use of consensus code and standards. The purpose of this study is to identify applicable codes and standards for the WP Emplacement Gantry ITS SSCs. Further, this study will form the basis for selection and the extent of applicability of each code and standard. This study is based on the design development completed for LA only. Accordingly, identification of ITS SSCs beyond those defined within the NSDB are based on designs that may be subject to further development during detail design. Furthermore, several design alternatives may still be under consideration to satisfy certain safety functions, and that final selection will not be determined until further design development has occurred. Therefore, for completeness, throughout this study alternative designs currently under considered will be discussed. Further, the results of this study will be subject to evaluation as part of a follow-on GAP analysis study. Based on the results of this study the GAP analysis will evaluate each code and standard to ensure each ITS performance requirement is fully satisfied. When a performance requirement is not fully satisfied a ''gap'' is highlighted. Thereafter, the study will identify supplemental requirements to augment the code or standard to meet performance requirements. Further, the GAP analysis will identify non-standard areas of the design that will be subject to a Development Plan. Non-standard components and non-standard design configurations are defined as areas of the design that do not follow standard industry practices or codes and standards. Whereby, performance confirmation cannot be
TBV 361 RESOLUTION ANALYSIS: EMPLACEMENT DRIFT ORIENTATION

International Nuclear Information System (INIS)

Lin, M.; Kicker, D.C.; Sellers, M.D.

1999-01-01

The purpose of this To Be Verified/To Be Determined (TBX) resolution analysis is to release ''To Be Verified'' (TBV)-361 related to the emplacement drift orientation. The system design criterion in ''Subsurface Facility System Description Document'' (CRWMS M andO 1998a, p.9) specifies that the emplacement drift orientation relative to the dominant joint orientations should be at least 30 degrees. The specific objectives for this analysis include the following: (1) Collect and evaluate key block data developed for the repository host horizon rock mass. (2) Assess the dominant joint orientations based on available fracture data. (3) Document the maximum block size as a function of drift orientation. (4) Assess the applicability of the drift orientation/joint orientation offset criterion in the ''Subsurface Facility System Description Document'' (CRWMS M andO 1998a, p.9). (5) Consider the effects of seepage on drift orientation. (6) Verify that the viability assessment (VA) drift orientation complies with the drift orientation/joint orientation offset criterion, or provide justifications and make recommendations for modifying the VA emplacement drift layout. In addition to providing direct support to the System Description Document (SDD), the release of TBV-361 will provide support to the Repository Subsurface Design Department. The results from this activity may also provide data and information needs to support the MGR Requirements Department, the MGR Safety Assurance Department, and the Performance Assessment Organization
Longevity of Emplacement Drift Ground Support Materials

International Nuclear Information System (INIS)

Tang, D.

2000-01-01

The purpose of this analysis is to evaluate the factors affecting the longevity of emplacement drift ground support materials and to develop a basis for selection of materials for ground support that will function throughout the preclosure period. The Development Plan (DP) for this analysis is given in CRWMS M and O (Civilian Radioactive Waste Management System Management and Operating Contractor) (1999a). The candidate materials for ground support are steel (carbon steel, ductile cast iron, galvanized steel, and stainless steel, etc.) and cement. Steel will mainly be used for steel sets, lagging, channels, rock bolts, and wire mesh. Cement usage is only considered in the case of grouted rock bolts. The candidate materials for the invert structure are steel and crushed rock ballast. The materials shall be evaluated for the repository emplacement drift environment under a specific thermal loading condition based on the proposed License Application Design Selection (LADS) design. The analysis consists of the following tasks: (1) Identify factors affecting the longevity of ground control materials for use in emplacement drifts. (2) Review existing documents concerning behavior of candidate ground control materials during the preclosure period. The major criteria to be considered for steel are mechanical and thermal properties, and durability, of which corrosion is the most important concern. (3) Evaluate the available results and develop recommendations for material(s) to be used
Generic repository concept for RBMK-1500 spent nuclear fuel disposal in crystalline rocks in Lithuania

International Nuclear Information System (INIS)

Poskas, P.; Brazauskaite, A.; Narkunas, E.; Smaizys, A.; Sirvydas, A.

2006-01-01

During 2002-2005 investigations on possibilities to dispose of spent nuclear fuel (SNF) in Lithuania were performed with support of Swedish experts. Disposal concept for RBMK-1500 SNF in crystalline rocks in Lithuania is based on Swedish KBS-3 concept with SNF emplacement into the copper canister with cast iron insert. The bentonite and its mixture with crushed rock are also foreseen as buffer and backfill material. In this paper modelling results on thermal, criticality and other important disposal characteristics for RBMK-1500 SNF fuel emplaced in copper canisters are presented. Based on thermal calculations, the distances between the canisters and between the tunnels were justified. Criticality calculations for the canister with fresh fuel with 2.8 % 235 U enrichment demonstrated that effective neutron multiplication factor k eff values are less than allowable value of 0.95. Dose calculations have shown that total equivalent dose rate from the canister with 50 years stored RBMK-1500 SNF is rather high and is defined mainly by the γ radiation. (author)
Estimated routine radiation doses to transportation workers in alternative spent-fuel transportation systems

International Nuclear Information System (INIS)

Schneider, K.J.; Smith, R.I.; Daling, P.M.; Ross, W.A.; McNair, G.W.

1988-01-01

The federal system for the management of spent fuel and high-level radioactive waste includes the acceptance by the US Department of Energy (DOE) of the spent fuel or waste loaded in casks at the reactor or other waste generators, its transportation to a repository, and its handling and final emplacement in the repository. The DOE plans to implement a transportation system that is safe, secure, efficient, and cost-effective and will meet applicable regulatory safety and security requirements. The DOE commissioned the Pacific Northwest Laboratory (PNL) to develop estimates of the routine radiation doses that would result from the operation of a system postulated using current designs and practices. From that evaluation, PNL identified activities/operations that result in the higher fraction of doses, proposed conceptual alternatives that would effectively reduce such exposures, and evaluated the cost-effectiveness of such alternatives. The study is one of a series used in making overall system design and operational decisions in the development of the DOE's spent-fuel/high-level waste transportation system. This paper contains the highlights from the PNL study of the estimated radiation doses to the transportation workers in a postulated reference transportation system and potential alternatives to that system
Solar energy emplacement developer

Science.gov (United States)

Mortensen, Michael; Sauls, Bob

1991-01-01

A preliminary design was developed for a Lunar Power System (LPS) composed of photovoltaic arrays and microwave reflectors fabricated from lunar materials. The LPS will collect solar energy on the surface of the Moon, transform it into microwave energy, and beam it back to Earth where it will be converted into usable energy. The Solar Energy Emplacement Developer (SEED) proposed will use a similar sort of solar energy collection and dispersement to power the systems that will construct the LPS.
Salt Repository Project waste emplacement mode decision paper: Revison 1

International Nuclear Information System (INIS)

1987-08-01

This paper provides a recommendation as to the mode of waste emplacement to be used as the current basis for site characterization activity for the Deaf Smith County, Texas, high level nuclear waste repository site. It also presents a plan for implementing the recommendation so as to provide a high level of confidence in the project's success. Since evaluations of high-level waste disposal in geologic repositories began in the 1950s, most studies emplacement in salt formations employed the vertical orientation for emplacing waste packages in boreholes in the floor of the underground facility. This orientation was used in trials at Project Salt Vault in the 1960s. The Waste Isolation Pilot Plant (WIPP) has recently settled on a combination of vertical and horizontal modes for various waste types. This paper analyzes the information available and develops a project position upon which to base current site characterization activities. The position recommended is that the SRP should continue to use the vertical waste emplacement mode as the reference design and to carry the horizontal mode as a ''passive'' alternative. This position was developed based upon the conclusions of a decision analysis, risk assessment, and cost/schedule impact assessment. 52 refs., 6 figs., 1 tab
Waste Isolation Pilot Plant in situ experimental program for HLW

International Nuclear Information System (INIS)

Molecke, M.A.

1977-01-01

The Waste Isolation Pilot Plant (WIPP) will be a facility to demonstrate the environmental and operational safety of storing radioactive wastes in a deep geologic bedded salt facility. The WIPP will be located in southeastern New Mexico, approximately 30 miles east of the city of Carlsbad. The major focus of the pilot plant operation involves ERDA defense related low and intermediate-level transuranic wastes. The scope of the project also specifically includes experimentation utilizing commercially generated high-level wastes, or alternatively, spent unreprocessed fuel elements. WIPP HLW experiments are being conducted in an inter-related laboratory, bench-scale, and in situ mode. This presentation focuses on the planned in situ experiments which, depending on the availability of commercially reprocessed waste plus delays in the construction schedule of the WIPP, will begin in approximately 1985. Such experiments are necessary to validate preceding laboratory results and to provide actual, total conditions of geologic storage which cannot be adequately simulated. One set of planned experiments involves emplacing bare HLW fragments into direct contact with the bedded salt environment. A second set utilizes full-size canisters of waste emplaced in the salt in the same manner as planned for a future HLW repository. The bare waste experiments will study in an accelerated manner waste-salt bed-brine interactions including matrix integrity/degradation, brine leaching, system chemistry, and potential radionuclide migration through the salt bed. Utilization of full-size canisters of HLW in situ permits us to demonstrate operational effectiveness and safety. Experiments will evaluate corrosion and compatibility interactions between the waste matrix, canister and overpack materials, getter materials, stored energy, waste buoyancy, etc. Using full size canisters also allows us to demonstrate engineered retrievability of wastes, if necessary, at the end of experimentation
Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

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

1994-01-01

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

International Nuclear Information System (INIS)

Griffiths, S.K.; Morrison, F.A. Jr.

1979-01-01

To determine the permeabilities of granular materials emplaced in vertical drill holes used for underground nuclear tests, an experiment at the USDOE Nevada Test Site (NTS) was conducted. As the hole is being filled, falling material increases pressure above and within the granular beds beneath. When the filling operation starts or stops, a transient pressure response occurs within the beds; measurements of this response in beds of various compositions were made. The permeabilities after emplacement were found by matching analytical predictions of the response to these data. This information is useful in assuring the containment of nuclear tests conducted in such drill holes
Spent Fuel Test - Climax: technical measurements. Interim report, fiscal year 1982

International Nuclear Information System (INIS)

Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

1983-02-01

The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted for the US Department of Energy (DOE) under the technical direction of the Lawrence Livermore National Laboratory (LLNL). Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April to May 1980, thus initiating a test with a planned 3- to 5-year fuel storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Three exchanges of spent fuel between the SFT-C and a surface storage facility furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and two previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 2-1/2 years of the test on more than 900 channels. Data continue to be acquired from the test. Some data are now available for analysis and are presented here. Highlights of activities this year include analysis of fracture data obtained during site characterization, laboratory studies of radiation effects and drilling damage in Climax granite, improved calculations of near-field heat transfer and thermomechanical response, a ventilation effects study, and further development of the data acquisition and management systems
Conceptual designs of automated systems for underground emplacement and retrieval of nuclear waste

International Nuclear Information System (INIS)

Slocum, A.H.; Hou, W.M.; Park, K.; Hochmuth, C.; Thurston, D.C.

1987-01-01

Current designs of underground nuclear waste repositories have not adequately addressed the possibility of automated, unmanned emplacement and retrieval. This report will present design methodologies for development of an automated system for underground emplacement of nuclear waste. By scaling generic issues to different repositories, it is shown that a two vehicle automated waste emplacement/retrieval system can be designed to operate in a fail-safe mode. Evaluation of cost at this time is not possible. Significant gains in worker safety, however, can be realized by minimizing the possibility of human exposure
Parameters and criteria influencing the selection of waste emplacement configurations in mined geologic repositories

International Nuclear Information System (INIS)

Bechthold, W.; Closs, K.D.; Papp, R.

1988-01-01

Reference concepts for repositories in deep geological formations have been developed in several countries. For these concepts, emplacement configurations vary within a wide range that comprises drift emplacement of unshielded or self-shielded packages and horizontal or vertical borehole emplacement. This is caused by different parameters, criteria, and criteria weighting factors. Examples for parameters are the country's nuclear power program and waste management policy, its geological situation, and safety requirements, examples for criteria and repository area requirements, expenditures of mining and drilling, and efforts for emplacement and, if required, retrieval. Due to the variety of these factors and their ranking in different countries, requirements for a safe, dependable and cost-effective disposal of radioactive waste can be met in various ways
Emplacement Gantry Gap Analysis Study

International Nuclear Information System (INIS)

Thornley, R.

2005-01-01

To date, the project has established important to safety (ITS) performance requirements for structures, systems, and components (SSCs) based on the identification and categorization of event sequences that may result in a radiological release. These performance requirements are defined within the ''Nuclear Safety Design Bases for License Application'' (NSDB) (BSC 2005 [DIRS 171512], Table A-11). Further, SSCs credited with performing safety functions are classified as ITS. In turn, assurance that these SSCs will perform as required is sought through the use of consensus codes and standards. This gap analysis is based on the design completed for license application only. Accordingly, identification of ITS SSCs beyond those defined within the NSDB are based on designs that may be subject to further development during detail design. Furthermore, several design alternatives may still be under consideration to satisfy certain safety functions, and final selection will not be determined until further design development has occurred. Therefore, for completeness, alternative designs currently under consideration will be discussed throughout this study. This gap analysis will evaluate each code and standard identified within the ''Emplacement Gantry ITS Standards Identification Study'' (BSC 2005 [DIRS 173586]) to ensure each ITS performance requirement is fully satisfied. When a performance requirement is not fully satisfied, a gap is highlighted. This study will identify requirements to supplement or augment the code or standard to meet performance requirements. Further, this gap analysis will identify nonstandard areas of the design that will be subject to a design development plan. Nonstandard components and nonstandard design configurations are defined as areas of the design that do not follow standard industry practices or codes and standards. Whereby, assurance that an SSC will perform as required may not be readily sought though the use of consensus standards. This
Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

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

1994-01-01

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

International Nuclear Information System (INIS)

Nair, B.R.; Bahorich, R.J.

1987-01-01

The current design concept for the disposal of nuclear high level waste packages in a repository in salt is based on the emplacement of individual packages in vertical boreholes in the underground mine floor. A key requirement is that the waste packages be capable of being retrieved during the last 26 years of the 76-year repository operating period. The unique design considerations relating to the retrieval of waste packages emplaced in bedded salt are presented in this paper. The information is based on the experience developed during the design of vertical emplacement and retrieval equipment in support of the Sandia Defense High Level Waste experiments at the Waste Isolation Pilot Plant. Also included are the impact of retrievability on the design of the equipment, the special salt cutting technology that was developed for this application, and a description of the equipment
Equipment for the emplacement of heat-producing waste in long horizontal boreholes

International Nuclear Information System (INIS)

Young, K.D.; Scully, L.W.; Fisk, A.; deBakker, P.; Friant, J.; Anderson, A.

1983-01-01

Emplacement of heat-producing waste in long horizontal holes may offer several technical and economic advantages over shallow vertical hole emplacement. Less of the host rock suffers damage as a result of drift construction; the heat from the waste can be isolated from the access drifts for long periods of time; and the amount of rock which must be excavated is much less than in traditional disposal scenarios. One of the major reasons that has been used to reject the long hole concept in the past and adhere to the shallow vertical hole concept is the equipment required to drill the holes and to emplace and retrieve the waste. Such equipment does not currently exist. It clearly is more difficult to drill a 600 to 1000 foot horizontal hole, possibly 3 to 4 feet in diameter, and place a canister of waste at the end of it than to drill a 30 foot vertical hole and lower the waste to the bottom. A liner, for emplacement hole stabilization, appears to be feasible by adapting existing technology for concrete slip forming or jacking in a steel liner. The conceptual design of the equipment to drill long horizontal holes, emplace waste and retrieve waste will be discussed. Various options in concept will be presented as well as their advantages and disadvantages. The operating scenario of the selected concept will be described as well as solutions to potential problems encountered
Equipment for the emplacement of heat-producing waste in long horizontal boreholes

International Nuclear Information System (INIS)

Young, K.D.; Fisk, A.; Friant, J.; Scully, L.W.

1983-01-01

Emplacement of heat-producing waste in long horizontal holes may offer several technical and economic advantages over shallow vertical hole emplacement. Less of the host rock suffers damage as a resul of drift construction; the heat from the waste can be isolated from the access drifts for long periods of time; and the amount of rock which must be excavated is much less than in traditional disposal scenarios. One of the major reasons that has been used to reject the long hole concept in the past and adhere to the shallow vertical hole concept is the equipment required to drill the holes and to emplace and retrieve the waste. Such equipment does not currently exist. It clearly is more difficult to drill a 600 to 100 foot horizontal hole, possibly 3 to 4 feet in diameter, and place a canister of waste at the end of it than to drill a 30 foot vertical hole and lower the waste to the bottom. A liner, for emplacement hole stabilization, appears to be feasible by adapting existing technology for concrete slip forming or jacking in a steel liner. The conceptual design of the equipment to drill long horizontal holes, emplace waste and retrieve waste is discussed. Various options in concept are presented as well as their advantages and disadvantages. The operating scenario of the selected concept is described as well as solutions to potential problems encountered

Strike-slip pull-apart process and emplacement of Xiangshan uranium-producing volcanic basin

International Nuclear Information System (INIS)

Qiu Aijin; Guo Lingzhi; Shu Liangshu

2001-01-01

Xiangshan volcanic basin is one of the famous uranium-producing volcanic basins in China. Emplacement mechanism of Xiangshan uranium-producing volcanic basin is discussed on the basis of the latest research achievements of deep geology in Xiangshan area and the theory of continental dynamics. The study shows that volcanic activity in Xiangshan volcanic basin may be divided into two cycles, and its emplacement is controlled by strike-ship pull-apart process originated from the deep regional faults. Volcanic apparatus in the first cycle was emplaced in EW-trending structure activated by clockwise strike-slipping of NE-trending deep fault, forming the EW-trending fissure-type volcanic effusion belt. Volcanic apparatus in the second cycle was emplaced at junction points of SN-trending pull-apart structure activated by sinistral strike-slipping of NE-trending deep faults and EW-trending basement faults causing the center-type volcanic magma effusion and extrusion. Moreover, the formation mechanism of large-rich uranium deposits is discussed as well
Investigations of possibilities to dispose of spent nuclear fuel in Lithuania: a model case. Volume 3, Generic Safety Assessment of Repository in Crystalline Rocks

International Nuclear Information System (INIS)

Motiejunas, S.; Poskas, P.

2005-01-01

In this Volume a generic safety assessment of the repository for spent nuclear fuel in crystalline rock in Lithuania is presented. Modeling of safety relevant radionuclide release from the defected canister and their transport through the near field and far field was performed. Doses to humans due to released radionuclides in the well water were calculated and compared with the dose restrictions existing in Lithuania. For this stage of generic safety assessment only two scenarios were chosen: base scenario and canister defect scenario. KBS-3 concept developed by SKB for disposal of spent nuclear fuel in Sweden was chosen as prototype for repository in crystalline basement in Lithuania. The KBS-3H design with horizontal canister emplacement is proposed as a reference design for Lithuania
Gamma-radiolysis of organic compounds and alpha-radiolysis of water

International Nuclear Information System (INIS)

Christensen, H.

1978-09-01

This KBS-report is a collection of five technical reports. Various radiolytic problems in connection with the disposal of high-active waste are dealt with. The English titles of the five reports are: 1) Radiolysis of fulvic acids; 2) Radiolysis of organic compounds in bentonite; 3) α-radiolysis of water during the disposal of fuel without reprocessing; 4) Radiolysis of water during the disposal of unreprocessed spent fuel. Oxidation by hydrogen peroxide; 5) Formation and decomposition of hydrogen peroxide by α-radiolysis. These reports (in Swedish) are attached as Appendices A-E. (author)
Areal thermal loading recommendations for nuclear waste repositories in salt

International Nuclear Information System (INIS)

Russell, J.E.

1979-06-01

This document gives a wider understanding of the history of the recommended thermal loadings in salt for both high-level waste (HLW) from fresh UO 2 -fueled, light-water reactors (LWR) with no recycle and spent unreprocessed fuel (SURF) from LWRs. Aspects of the current recommendations that need further study are identified. Finally, an interim set of design thermal-loading recommendations are given that have a common rationale of satisfying performance limits within our current state of knowledge. These recommendations are made on a generic rather than a site-specific basis. 11 figures, 5 tables
STATE OF THE ART OF DRILLING LARGE DIAMETER BOREHOLES FOR DEPOSITION OF HIGH LEVEL WASTE AND SPENT NUCLEAR FUEL

Directory of Open Access Journals (Sweden)

Trpimir Kujundžić

2012-07-01

Full Text Available Deep geological disposal is internationally recognized as the safest and most sustainable option for the long-term management of high-level radioactive waste. Mainly, clay rock, salt rock and crystalline rock are being considered as possible host rocks. Different geological environment in different countries led to the various repository concepts. Main feature of the most matured repository concept is that canisters with spent nuclear fuel are emplaced in vertical or horizontal large diameter deposition holes. Drilling technology of the deposition holes depends on repository concept and geological and geomechanical characteristics of the rock. The deposition holes are mechanically excavated since drill & blast is not a possible method due to requirements on final geometry like surface roughness etc. Different methods of drilling large diameter boreholes for deposition of high-level waste and spent nuclear fuel are described. Comparison of methods is made considering performance and particularities in technology.
Nuclear criticality safety analysis of a spent fuel waste package in a tuff repository

International Nuclear Information System (INIS)

Weren, B.H.; Capo, M.A.; O'Neal, W.C.

1983-12-01

An assessment has been performed of the criticality potential associated with the disposal of spent fuel in a tuff geology above the water table. Eleven potential configurations were defined which cover a vast range of geometries and conditions from the nominal configuration at emplacement to a hypothetical configuration thousands of years after emplacement in which the structure is gone, the fuel pellets disintegrated and the borehole flooded. Of these eleven configurations, four have been evaluated at this time. The results of this evaluation indicate that even with very conservative assumptions (4.5 w/o fresh fuel), criticality is not a problem for the nominal configuration either dry or fully flooded. In the cases where the condition of the waste package is assumed to have severely deteriorated, over long times, calculations were performed with less conservative assumptions (depleted fuel). An assessment of these calculations indicates that criticality safety could be demonstrated if the depletion of the fissile inventory during fuel irradiation is taken into account. A detailed discussion of the calculations performed is presented in this report. Also included are a description of the configurations which were considered, the analytical methods and models used, and a discussion of additional related work which should be performed. 15 references, 11 figures, 8 tables
Effects of gaseous radioactive nuclides on the design and operation of repositories for spent LWR fuel in rock salt

International Nuclear Information System (INIS)

Jenks, G.H.

1979-12-01

Information relating to the identities and amounts of gaseous radionuclides present in spent LWR fuel and to their release from canistered spent fuel under plausible storage and disposal conditions was assembled, reviewed, and analyzed. Information was also reviewed and analyzed on several other subjects that relate to the integrity of the carbon steel canister in which the spent fuel is to be encapsulated and to the expected rates of transfer of gaseous radionuclides through crushed salt backfill within a disposal room in a reference repository in rock salt. The advantages and disadvantages were considered for several different canister-backfill materials, and recommendations were made regarding preferred materials. Other recommendations relate to encapsulation procedures and specifications and to needs for additional experimental studies. The objective of this work was to provide reference information, conclusions, and recommendations that could be used to establish design and operating conditions and procedures for a bedded salt repository for spent LWR fuel and that could also be used to help evaluate the safety of the repository. The results of this work will also generally apply to spent fuel repositories in domal salt. However, because the domal salt may have little or no brine inclusions within it, there may be little or no possibility that brine will migrate into open spaces around an emplaced canister. Addordingly, some of the concerns that result from the possible occurrence of brine migration in bedded salt may be of no importance in domal salt
Thermally-assisted Magma Emplacement Explains Restless Calderas.

Science.gov (United States)

Amoruso, Antonella; Crescentini, Luca; D'Antonio, Massimo; Acocella, Valerio

2017-08-11

Many calderas show repeated unrest over centuries. Though probably induced by magma, this unique behaviour is not understood and its dynamics remains elusive. To better understand these restless calderas, we interpret deformation data and build thermal models of Campi Flegrei caldera, Italy. Campi Flegrei experienced at least 4 major unrest episodes in the last decades. Our results indicate that the inflation and deflation of magmatic sources at the same location explain most deformation, at least since the build-up of the last 1538 AD eruption. However, such a repeated magma emplacement requires a persistently hot crust. Our thermal models show that this repeated emplacement was assisted by the thermal anomaly created by magma that was intruded at shallow depth ~3 ka before the last eruption. This may explain the persistence of the magmatic sources promoting the restless behaviour of the Campi Flegrei caldera; moreover, it explains the crystallization, re-melting and mixing among compositionally distinct magmas recorded in young volcanic rocks. Our model of thermally-assisted unrest may have a wider applicability, possibly explaining also the dynamics of other restless calderas.
Emplacement feasibility of a multi-tier, expanded capacity repository at Yucca Mountain, Nevada USA

International Nuclear Information System (INIS)

Apted, Michael; Kessler, John; Fairhurst, Charles

2008-01-01

A geological repository at Yucca Mountain has been proposed for the disposal of spent fuel from the US commercial reactors and other radioactive waste. A legislative capacity of 70,000 MTHM has been set by the Nuclear Waste Policy Act of 1982, including 63,000 MTHM of commercial spent nuclear fuel (CSNF), the projected amount of CSNF that will be produced by about 2014. Policy issues remain as to how to handle waste that is generated beyond 2014 from a growing nuclear industry in the US. The Electric Power Research Institute (EPRI) is independently evaluating the technical, rather than legislative, limit of CSNF that could be safely disposed at Yucca Mountain. Geological, thermal management, safety and cost factors have been recently evaluated by EPRI (2006; 2007) for grouped emplacement drifts and/or a multi-tier repository. EPRI's evaluation of emplacement feasibility for a multi-tier concept is described here. Expanded capacity concepts as envisioned for Yucca Mountain (EPRI, 2006; 2007) assume excavation of one or two additional levels of drifts parallel to or above and/or below the original drift excavations. For the latter multi-tier concept each 'tier' or 'level' would essentially replicate the original layer with a 30-m separation between tiers. This arrangement essentially doubles or triples the capacity of the repository for a two- or three-tier design, respectively. The main issues that affect the feasibility of expanded capacity design are; (i) ventilation requirements; (ii) radiation hazards; (iii) thermal and thermo-mechanical constraints. (i)Ventilation: The repository design involves waste packages mounted in close proximity to each other in 600-m long drifts that remain open and actively ventilated for at least 50-100 years. Analyses,conservatively assuming that all three repository levels operate simultaneously, indicate no technological obstacles in meeting ventilation requirements for sustained simultaneous operation ba sed on current industrial
Plan for safety case of spent fuel repository at Olkiluoto

International Nuclear Information System (INIS)

Vieno, T.; Ikonen, A.T.K.

2005-02-01

Posiva aims to present the Safety Case supporting the construction license application of the spent fuel repository at Olkiluoto by 2012. An outline and preliminary assessments will be presented in 2009. Interim reporting and an update of the Safety Case plan will be presented in 2006, as required by the authorities. The KBS-3 disposal concept aims at long-term isolation and containment of spent fuel assemblies in durable copper-iron canisters emplaced in a repository to be constructed at a depth between 400 and 600 metres in crystalline bedrock. By 2012, studies on the KBS-3 disposal concept and site investigations at Olkiluoto will have been continued over about thirty years. The construction of an underground rock characterisation facility (called ONKALO) was started in June 2004. The investigations are carried out in close cooperation with the Swedish SKB developing and assessing the same disposal concept at candidate sites, resembling Olkiluoto, at the other side of the Baltic Sea. A safety case is the synthesis of evidence, analyses and arguments that quantify and substantiate the safety, and the level of expert confidence in the safety, of a planned repository. Posiva's Safety Case will be organised in a portfolio including ten main reports, which will be periodically updated according the overall schedule presented in the plan. The Site report describing the present state and past evolution of the Olkiluoto site, as well as the disturbances caused by the construction of ONKALO and the first stage of the repository, forms the geoscientific basis of the Safety Case. The engineering basis is provided by the reports on the Characteristics of spent fuel, Canister design, and Repository design. The Process report containing descriptions and analyses of features, events and processes potentially affecting the disposal system, and the report on the Evolution of site and repository form the scientific basis of the Safety Case. The latter report will describe and
Possibility of multiple temperature maxima in geologic repositories for spent fuel from nuclear reactors

International Nuclear Information System (INIS)

Beyerlein, S.W.; Claiborne, H.C.

1980-01-01

Heat transfer studies show that two temperature maxima at the disposal horizon could be experienced in CANDU spent fuel repositories - one at about 60 years and another slightly higher one at 13,000 years. Because CANDU spent fuels display a monotonically decreasing heat generation rate, it is not immediately obvious why this behavior should occur. This report investigates this behavior, confirms the Canadian results, demonstrates that the double peak phenomenon is due to the presence of the right mixture of short- and long-lived nuclides in the fuel, and concludes that the 13,000-year maximum is largely an artifact of the infinite or very large plane source model. When more realistic repository geometries are used, the second peak disappears for repository sizes less than about 1 km 2 . Over the long term, radial and surface heat transfer causes the thermal history of the disposal region to deviate from that predicted by infinite plane (or large finite) source models by reducing the magnitude of the second peak. Beyond a 1000-year time horizon, care should be exercised in modeling spent fuel repositories to include the proper boundary conditions. For the first few centuries after emplacement, however, the infinite source model is consistent with the finite disk source model as well as with arrays of spherical and point sources. The second temperature peak can be avoided by restricting the size of the repository and/or partitioning out the long-lived components of the fuel. When spent fuel from PWRs was examined for multiple temperature maxima, only one peak was found, even for the infinite plane source model
Magma emplacement in 3D

Science.gov (United States)

Gorczyk, W.; Vogt, K.

2017-12-01

Magma intrusion is a major material transfer process in Earth's continental crust. Yet, the mechanical behavior of the intruding magma and its host are a matter of debate. In this study, we present a series of numerical thermo-mechanical experiments on mafic magma emplacement in 3D.In our model, we place the magmatic source region (40 km diameter) at the base of the mantle lithosphere and connect it to the crust by a 3 km wide channel, which may have evolved at early stages of magmatism during rapid ascent of hot magmatic fluids/melts. Our results demonstrate continental crustal response due to magma intrusion. We observe change in intrusion geometries between dikes, cone-sheets, sills, plutons, ponds, funnels, finger-shaped and stock-like intrusions as well as injection time. The rheology and temperature of the host-rock are the main controlling factors in the transition between these different modes of intrusion. Viscous deformation in the warm and deep crust favours host rock displacement and magma pools along the crust-mantle boundary forming deep-seated plutons or magma ponds in the lower to middle-crust. Brittle deformation in the cool and shallow crust induces cone-shaped fractures in the host rock and enables emplacement of finger- or stock-like intrusions at shallow or intermediate depth. A combination of viscous and brittle deformation forms funnel-shaped intrusions in the middle-crust. Low-density source magma results in T-shaped intrusions in cross-section with magma sheets at the surface.
Sediment mechanical response due to emplacement of a waste canister

International Nuclear Information System (INIS)

Karnes, C.H.; Dawson, P.R.; Silva, A.J.; Brown, W.T.

1980-01-01

Preliminary studies have been conducted to determine the interaction between a waste canister and seabed sediment during and after emplacement. Empirical and approximate methods for determining the depth reached by a freefall penetrator indicate that a boosted penetrator emplacement method may be necessary. Hole closure is necessary, but has not been verified because calculations and laboratory experiments show sensitivity to boundary conditions which control the degree of dynamic hole closure. Laboratory studies show that closure will take place by creep deformation but closure times in seabed environments are uncertain. For assumed thermomechanical properties of sediments, it is shown that a heat generating waste canister will probably not move a significant distancce during the heat generation period
From steep feeders to tabular plutons - Emplacement controls of syntectonic granitoid plutons in the Damara Belt, Namibia

Science.gov (United States)

Hall, Duncan; Kisters, Alexander

2016-01-01

Granitoid plutons in the deeply eroded south Central Zone of the Damara Belt in Namibia commonly show tabular geometries and pronounced stratigraphic controls on their emplacement. Subhorizontal, sheet-like pluton geometries record emplacement during regional subhorizontal shortening, but the intrusion of spatially and temporally closely-related granitoid plutons at different structural levels and in distinct structural settings suggests independent controls on their levels of emplacement. We describe and evaluate the controls on the loci of the dyke-to-sill transition that initiated the emplacement of three syntectonic (560-530 Ma) plutons in the basement-cover stratigraphy of the Erongo region. Intrusive relationships highlight the significance of (1) rigidity anisotropies associated with competent sedimentary packages or pre-existing subhorizontal granite sheets and (2) rheological anisotropies associated with the presence of thick ductile marble horizons. These mechanical anisotropies may lead to the initial deflection of steep feeder conduits as well as subsequent pluton assembly by the repeated underaccretion of later magma batches. The upward displacement of regional isotherms due to the heat advection associated with granite emplacement is likely to have a profound effect on the mechanical stratification of the upper crust and, consequently, on the level at which granitoid pluton emplacement is initiated. In this way, pluton emplacement at progressively shallower crustal depths may have resulted in the unusually high apparent geothermal gradients recorded in the upper crustal levels of the Damara Belt during its later evolution.
Conditioning of spent fuel assemblies from the Rossendorf RFR research reactor in transport and storage containers of the type CASTOR MTR 2

International Nuclear Information System (INIS)

Schneider, B.; Hofmann, G.

1994-09-01

Most of the spent fuel assemblies are temporarily stored in the flooded fuel ponds AB 1 and AB 2 of the RFR, and some are still in the reactor core. The conditioning task described here is part of the RFR spent fuel management concept and covers the safe emplacement of the spent fuel elements in the CASTOR MTR 2 shipping containers and the sealing of the containers in compliance with the nuclear licence issued for the conditioning task. The transfer of the spent fuel assemblies from the present wet storage conditions to the dry storage conditions in the CASTOR MTR 2 containers is done by a mobile manipulation equipment consisting essentially of the transfer sluice gate and a transfer container. Subsequent to conditioning, the shipping containers are to be transported to a licensed intermediate storage facility to await their transport to a national radwaste repository. The technical handling tools for the transfer and manipulation are briefly described, as well as the process steps involved, putting emphasis on the detailed description of processes and the accompanying time frame, so that the conditioning task can be incorporated into the work plan of the entire project. The report further presents the EDP concept established for the task, including the required data archivation and documentation. (orig.) [de
On the environmental impact of a repository for spent nuclear fuel

International Nuclear Information System (INIS)

Brotzen, O.

1983-04-01

A worst case estimate is presented of the contamination of groundwater discharging at the surface. The contamination originates from uranium and actinides in spent nuclear fuel emplaced at ca 500 m depth in fractured, water-saturated, crystalline rock. No engineered barriers are considered, the fuel is free to interact with ground-water and the host rock. Redox processes in a radiolysis zone, and in interaction with the rock, precipitations and absorptions are considered. The contaminated water is assumed to reach the surface through a highly water-conducting fracture zone. The contamination would then take more than 300000 years to reach the surface, and would mainly consist of uranium, actinides would not exceed their ratio to uranium in the fuel. The fuel uranium content in surfacing ground-water would be of the order of one percent of the average surface well content in Sweden. (P.Aa.)
Systems study on engineered barriers: barrier performance analysis

International Nuclear Information System (INIS)

Stula, R.T.; Albert, T.E.; Kirstein, B.E.; Lester, D.H.

1980-09-01

A performance assessment model for multiple barrier packages containing unreprocessed spent fuel has been modified and applied to several package designs. The objective of the study was to develop information to be used in programmatic decision making concerning engineered barrier package design and development. The assessment model, BARIER, was developed in previous tasks of the System Study on Engineered Barriers (SSEB). The new version discussed in this report contains a refined and expanded corrosion rate data base which includes pitting, crack growth, and graphitization as well as bulk corrosion. Corrosion rates for oxic and anoxic conditions at each of the two temperature ranges are supplied. Other improvements include a rigorous treatment of radionuclide release after package failure which includes resistance of damaged barriers and backfill, refined temperature calculations that account for convection and radiation, a subroutine to calculate nuclear gamma radiation field at each barrier surface, refined stress calculations with reduced conservatism and various coding improvements to improve running time and core usage. This report also contains discussion of alternative scenarios to the assumed flooded repository as well as the impact of water exclusion backfills. The model was used to assess post repository closure performance for several designs which were all variation of basic designs from the Spent Unreprocessed Fuel (SURF) program. Many designs were found to delay the onset of leaching by at least a few hundreds of years in all geologic media. Long delay times for radionuclide release were found for packages with a few inches of sorption backfill. Release of uranium, plutonium, and americium was assessed
Systems study on engineered barriers: barrier performance analysis

Energy Technology Data Exchange (ETDEWEB)

Stula, R.T.; Albert, T.E.; Kirstein, B.E.; Lester, D.H.

1980-09-01

A performance assessment model for multiple barrier packages containing unreprocessed spent fuel has been modified and applied to several package designs. The objective of the study was to develop information to be used in programmatic decision making concerning engineered barrier package design and development. The assessment model, BARIER, was developed in previous tasks of the System Study on Engineered Barriers (SSEB). The new version discussed in this report contains a refined and expanded corrosion rate data base which includes pitting, crack growth, and graphitization as well as bulk corrosion. Corrosion rates for oxic and anoxic conditions at each of the two temperature ranges are supplied. Other improvements include a rigorous treatment of radionuclide release after package failure which includes resistance of damaged barriers and backfill, refined temperature calculations that account for convection and radiation, a subroutine to calculate nuclear gamma radiation field at each barrier surface, refined stress calculations with reduced conservatism and various coding improvements to improve running time and core usage. This report also contains discussion of alternative scenarios to the assumed flooded repository as well as the impact of water exclusion backfills. The model was used to assess post repository closure performance for several designs which were all variation of basic designs from the Spent Unreprocessed Fuel (SURF) program. Many designs were found to delay the onset of leaching by at least a few hundreds of years in all geologic media. Long delay times for radionuclide release were found for packages with a few inches of sorption backfill. Release of uranium, plutonium, and americium was assessed.
Mineralogic and petrologic investigation of pre-test core samples from the spent fuel test-climax

International Nuclear Information System (INIS)

Ryerson, F.J.; Qualheim, B.J.

1983-12-01

Pre-test samples obtained from just inside the perimeter of the canister emplacement holes of the Spent Fuel Test-Climax have been characterized by petrographic and microanalytical techniques. The primary quartz monzonite has undergone various degrees of hydrothermal alteration as a result of natural processes. Alteration is most apparent on primary plagioclase and biotite. The most common secondary phases on plagioclase are muscovite and calcite, while the most common secondary phases on biotite are epidote and chlorite. The major alteration zones encountered are localized along filled fractures, i.e. veins. The thickness and mineralogy of the alteration zones can be correlated with the vein mineralogy, becoming wider and more complex mineralogically when the veins contain calcite. 7 references, 10 figures, 4 tables
Thermally-assisted Magma Emplacement Explains Restless Calderas

Science.gov (United States)

Amoruso, A.; Crescentini, L.; D'Antonio, M.; Acocella, V.

2017-12-01

Many calderas show repeated unrest over centuries. Though probably induced by magma, this unique behaviour is not understood and its dynamics remains elusive. To better understand these restless calderas, we interpret deformation data and build thermal models of Campi Flegrei, Italy, which is the best-known, yet most dangerous calderas, lying to the west of Naples and restless since the 1950s at least.Our elaboration of the geodetic data indicates that the inflation and deflation of magmatic sources at the same location explain most deformation, at least since the build-up of the last 1538 AD eruption. However, such a repeated magma emplacement requires a persistently hot crust.Our thermal models show that the repeated emplacement was assisted by the thermal anomaly created by magma that was intruded at shallow depth 3 ka before the last eruption and, in turn, contributed to maintain the thermal anomaly itself. This may explain the persistence of the magmatic sources promoting the restless behaviour of the Campi Flegrei caldera; moreover, it explains the crystallization, re-melting and mixing among compositionally distinct magmas recorded in young volcanic rocks.Available information at other calderas highlights similarities to Campi Flegrei, in the pattern and cause of unrest. All monitored restless calderas have either geodetically (Yellowstone, Aira Iwo-Jima, Askja, Fernandina and, partly, Long Valley) or geophysically (Rabaul, Okmok) detected sill-like intrusions inducing repeated unrest. Some calderas (Yellowstone, Long Valley) also show stable deformation pattern, where inflation insists on and mimics the resurgence uplift. The common existence of sill-like sources, also responsible for stable deformation patterns, in restless calderas suggests close similarities to Campi Flegrei. This suggests a wider applicability of our model of thermally-assisted sill emplacement, to be tested by future studies to better understand not only the dynamics of restless

Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

Energy Technology Data Exchange (ETDEWEB)

Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

1982-08-01

The thermal effects associated with the emplacement of aged radioactive wastes in a geologic repository were studied, with emphasis on the following subjects: the waste characteristics, repository structure, and rock properties controlling the thermally induced effects; the current knowledge of the thermal, thermomechanical, and thermohydrologic impacts, determined mainly on the basis of previous studies that assume 10-year-old wastes; the thermal criteria used to determine the repository waste loading densities; and the technical advantages and disadvantages of surface cooling of the wastes prior to disposal as a means of mitigating the thermal impacts. The waste loading densities determined by repository designs for 10-year-old wastes are extended to older wastes using the near-field thermomechanical criteria based on room stability considerations. Also discussed are the effects of long surface cooling periods determined on the basis of far-field thermomechanical and thermohydrologic considerations. The extension of the surface cooling period from 10 years to longer periods can lower the near-field thermal impact but have only modest long-term effects for spent fuel. More significant long-term effects can be achieved by surface cooling of reprocessed high-level waste.
Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

International Nuclear Information System (INIS)

Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

1982-08-01

The thermal effects associated with the emplacement of aged radioactive wastes in a geologic repository were studied, with emphasis on the following subjects: the waste characteristics, repository structure, and rock properties controlling the thermally induced effects; the current knowledge of the thermal, thermomechanical, and thermohydrologic impacts, determined mainly on the basis of previous studies that assume 10-year-old wastes; the thermal criteria used to determine the repository waste loading densities; and the technical advantages and disadvantages of surface cooling of the wastes prior to disposal as a means of mitigating the thermal impacts. The waste loading densities determined by repository designs for 10-year-old wastes are extended to older wastes using the near-field thermomechanical criteria based on room stability considerations. Also discussed are the effects of long surface cooling periods determined on the basis of far-field thermomechanical and thermohydrologic considerations. The extension of the surface cooling period from 10 years to longer periods can lower the near-field thermal impact but have only modest long-term effects for spent fuel. More significant long-term effects can be achieved by surface cooling of reprocessed high-level waste
Influence of heat flow on drift closure during Climax granite spent-fuel test: measurements and calculations

International Nuclear Information System (INIS)

Butkovich, T.R.; Yow, J.L. Jr.; Montan, D.N.

1982-01-01

Measurements of drift closure caused by the thermal load have been made routinely during the Spent Fuel Test in Climax granite since about six weeks after emplacement of the fuel. Horizontal and vertical closure was measured with a manually operated tape extensometer at various locations along the lengths of the drifts. Average closures ranged from 0 to 0.6 mm, horizontal and vertical, out to about 2.2 years since the start of the test. At the same time, displacements from the thermal loads were measured with rod extensometers emplaced to measure relative displacements between hole collars and anchor points in holes drilled from two parallel heater drifts. These data are compared with thermo-elastic finite element calculations which utilized measured properties of the Climax granite. The calculations show that more than half of the closures occur between fuel installation and the first closure measurement. The comparisons show that the results track each other, in that where closure followed by dilation is measured, the calculations also show this effect. The agreement is excellent, considering the averaged measured closures remain within 30% of the total calculated drift closures and the extremely small magnitude of the relative displacements (0.5 mm), measured or calculated
Description of recipient areas related to final storage of unreprocessed spent nuclear fuel

International Nuclear Information System (INIS)

Sundblad, B.; Bergstroem, U.

1983-02-01

A comprehensive study of recipient areas in Fjaellveden, Voxnan, Gideaa and Kamlungekoelen is accomplished. Besides general conditions in Finnsjoen and Sternoe are discussed. The recipient areas are defined and their climate, hydrology, bedrock, soil, vegetation, land use and yield from arable land are described as well as the yield of fish for the surface water of interest. The potential exposure pathways and model system at the different areas are defined. Long-term variations of geology, climate, hydrology, land-use, acidification and evolution are described. The possible development of the recipient areas is also discussed. (Authors)
Waste Handling and Emplacement Options for Disposal of Radioactive Waste in Deep Boreholes.

Energy Technology Data Exchange (ETDEWEB)

Cochran, John R.; Hardin, Ernest

2015-11-01

Traditional methods cannot be used to handle and emplace radioactive wastes in boreholes up to 16,400 feet (5 km) deep for disposal. This paper describes three systems that can be used for handling and emplacing waste packages in deep borehole: (1) a 2011 reference design that is based on a previous study by Woodward–Clyde in 1983 in which waste packages are assembled into “strings” and lowered using drill pipe; (2) an updated version of the 2011 reference design; and (3) a new concept in which individual waste packages would be lowered to depth using a wireline. Emplacement on coiled tubing was also considered, but not developed in detail. The systems described here are currently designed for U.S. Department of Energy-owned high-level waste (HLW) including the Cesium- 137/Strontium-90 capsules from the Hanford Facility and bulk granular HLW from fuel processing in Idaho.
Origin of elevated water levels encountered in Pahute Mesa emplacement boreholes: Preliminary investigations

International Nuclear Information System (INIS)

Brikowski, T.; Chapman, J.; Lyles, B.; Hokett, S.

1993-11-01

The presence of standing water well above the predicted water table in emplacement boreholes on Pahute Mesa has been a recurring phenomenon at the Nevada Test Site (NTS). If these levels represent naturally perched aquifers, they may indicate a radionuclide migration hazard. In any case, they can pose engineering problems in the performance of underground nuclear tests. The origin of these elevated waters is uncertain. Large volumes of water are introduced during emplacement drilling, providing ample source for artificially perched water, yet elevated water levels can remain constant for years, suggesting a natural origin instead. In an effort to address the issue of unexpected standing water in emplacement boreholes, three different sites were investigated in Area 19 on Pahute Mesa by Desert Research Institute (DRI) staff from 1990-93. These sites were U-19az, U-19ba, and U-19bh. As of this writing, U-19bh remains available for access; however, nuclear tests were conducted at the former two locations subsequent to this investigations. The experiments are discussed in chronological order. Taken together, the experiments indicate that standing water in Pahute Mesa emplacement holes originates from the drainage of small-volume naturally perched zones. In the final study, the fluids used during drilling of the bottom 100 m of emplacement borehole U-19bh were labeled with a chemical tracer. After hole completion, water level rose in the borehole, while tracer concentration decreased. In fact, total mass of tracer in the borehole remained constant, while water levels rose. After water levels stabilized in this hole, no change in tracer mass was observed over two years, indicating that no movement of water out of the borehole is taking place (as at U- 19ba). Continued labeling tests of standing water are recommended to confirm the conclusions made here, and to establish their validity throughout Pahute Mesa
Eruption and emplacement dynamics of a thick trachytic lava flow of the Sancy volcano (France)

Science.gov (United States)

Latutrie, Benjamin; Harris, Andrew; Médard, Etienne; Gurioli, Lucia

2017-01-01

A 70-m-thick, 2200-m-long (51 × 106 m3) trachytic lava flow unit underlies the Puy de Cliergue (Mt. Dore, France). Excellent exposure along a 400-m-long and 60- to 85-m-high section allows the flow interior to be accessed on two sides of a glacial valley that cuts through the unit. We completed an integrated morphological, structural, textural, and chemical analysis of the unit to gain insights into eruption and flow processes during emplacement of this thick silicic lava flow, so as to elucidate the chamber and flow dynamic processed that operate during the emplacement of such systems. The unit is characterized by an inverse chemical stratification, where there is primitive lava beneath the evolved lava. The interior is plug dominated with a thin basal shear zone overlying a thick basal breccia, with ramping affecting the entire flow thickness. To understand these characteristics, we propose an eruption model that first involves processes operating in the magma chamber whereby a primitive melt is injected into an evolved magma to create a mixed zone at the chamber base. The eruption triggered by this event first emplaced a trachytic dome, into which banded lava from the chamber base was injected. Subsequent endogenous dome growth led to flow down the shallow slope to the east on which the highly viscous (1012 Pa s) coulée was emplaced. The flow likely moved extremely slowly, being emplaced over a period of 4-10 years in a glacial manner, where a thick (>60-m) plug slid over a thin (5-m-thick) basal shear zone. Excellent exposure means that the Puy de Cliergue complex can be viewed as a case type location for understanding and defining the eruption and emplacement of thick, high-viscosity, silicic lava flow systems.
EMPLACEMENT DRIFT ISOLATION DOOR CONTROL SYSTEM

International Nuclear Information System (INIS)

N.T. Raczka

1998-01-01

The purpose of this analysis is to review and refine key design concepts related to the control system presently under consideration for remotely operating the emplacement drift isolation doors at the potential subsurface nuclear waste repository at Yucca Mountain. This analysis will discuss the key design concepts of the control system that may be utilized for remotely monitoring, opening, and closing the emplacement drift isolation doors. The scope and primary objectives of this analysis are to: (1) Discuss the purpose and function of the isolation doors (Presented in Section 7.1). (2) Review the construction of the isolation door and other physical characteristics of the doors that the control system will interface with (Presented in Section 7.2). (3) Discuss monitoring and controlling the operation of the isolation doors with a digital control system (either a Programmable Logic Controller (PLC) system or a Distributed Control System (DCS)) (Presented in Section 7.3). (4) Discuss how all isolation doors can be monitored and controlled from a subsurface central control center (Presented in Section 7.4). This analysis will focus on the development of input/output (I/O) counts including the types of I/O, redundancy and fault tolerance considerations, and processor requirements for the isolation door control system. Attention will be placed on operability, maintainability, and reliability issues for the system operating in the subsurface environment with exposure to high temperatures and radiation
Gamma irradiation tests of concrete material recommended for storage casks of spent nuclear fuel arising from Cernavoda NPP

International Nuclear Information System (INIS)

Dulama, M.; Deneanu, N.; Dulama, C.; Baboescu, E.

2001-01-01

Considerable effort is being devoted to the Romania's Nuclear Spent Fuel and Waste Management R and D Program to develop engineered barriers for the containment of nuclear fuel waste under conditions of deep geological disposal. Engineering practice suggests that the concrete should fulfil the requirements of long term physical stability and resistance to radiation. With an appropriate system of metal reinforcement, it should be possible to obtain the tensile and impact strength required, avoiding the risk of mechanical damage during handling and emplacement. In accordance with the concept developed by CITON-Bucharest, presently, the dry storage of spent nuclear fuel is thought by two choices: - The alternative of dry storage type MMB3; - The alternative of dry storage type TRANSTOR. By using ORIGEN and PELSHIE computer codes, we evaluated the gamma radiation dose absorbed by the concrete walls of the storage vault both in MMB3 and in TRANSTOR designing variants. The irradiation tests were performed at the Gamma Irradiation Facility of the Institute for Nuclear Research. (authors)
Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

International Nuclear Information System (INIS)

Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

2008-12-01

The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments that lie outside the
Uncertainty and sensitivity results for pre-waste-emplacement groundwater travel time

International Nuclear Information System (INIS)

Kaplan, P.G.

1992-01-01

In this paper uncertainty and sensitivity analyses for pre-waste-emplacement groundwater travel time conducted. Although preliminary, a numbed of interesting results were obtained. Uncertainty in the ground water travel time statistics, as measured by the coefficient of variation, increases and then decrease as the modeled system transitions from matrix-dominated to fracture-dominated flow. The uncertainty analysis also suggests that the median, as opposed to the mean, may be a better indicator of performance with respect to the regulatory criterion. The sensitivity analysis shows a strong correlation between an effective fracture property, fracture porosity, and failure to meet the regulatory pre-waste-emplacement groundwater travel time criterion of 1,000 years
Waste Handling and Emplacement Options for Disposal of Radioactive Waste in Deep Boreholes

International Nuclear Information System (INIS)

Cochran, John R.; Hardin, Ernest

2015-01-01

Traditional methods cannot be used to handle and emplace radioactive wastes in boreholes up to 16,400 feet (5 km) deep for disposal. This paper describes three systems that can be used for handling and emplacing waste packages in deep borehole: (1) a 2011 reference design that is based on a previous study by Woodward-Clyde in 1983 in which waste packages are assembled into ''strings'' and lowered using drill pipe; (2) an updated version of the 2011 reference design; and (3) a new concept in which individual waste packages would be lowered to depth using a wireline. Emplacement on coiled tubing was also considered, but not developed in detail. The systems described here are currently designed for U.S. Department of Energy-owned high-level waste (HLW) including the Cesium- 137/Strontium-90 capsules from the Hanford Facility and bulk granular HLW from fuel processing in Idaho.
Improvements relating to apparatus for emplacing or replacing a gaiter

International Nuclear Information System (INIS)

Chilvers, D.; Morrison, D.M.

1982-01-01

The invention relates to apparatus for emplacing or replacing a gaiter of the type which may be used in sealing the hot side of an isolation enclosure which is associated with a master-slave manipulator. (author)
Architecture and emplacement of flood basalt flow fields: case studies from the Columbia River Basalt Group, NW USA

Science.gov (United States)

Vye-Brown, C.; Self, S.; Barry, T. L.

2013-03-01

The physical features and morphologies of collections of lava bodies emplaced during single eruptions (known as flow fields) can be used to understand flood basalt emplacement mechanisms. Characteristics and internal features of lava lobes and whole flow field morphologies result from the forward propagation, radial spread, and cooling of individual lobes and are used as a tool to understand the architecture of extensive flood basalt lavas. The features of three flood basalt flow fields from the Columbia River Basalt Group are presented, including the Palouse Falls flow field, a small (8,890 km2, ˜190 km3) unit by common flood basalt proportions, and visualized in three dimensions. The architecture of the Palouse Falls flow field is compared to the complex Ginkgo and more extensive Sand Hollow flow fields to investigate the degree to which simple emplacement models represent the style, as well as the spatial and temporal developments, of flow fields. Evidence from each flow field supports emplacement by inflation as the predominant mechanism producing thick lobes. Inflation enables existing lobes to transmit lava to form new lobes, thus extending the advance and spread of lava flow fields. Minimum emplacement timescales calculated for each flow field are 19.3 years for Palouse Falls, 8.3 years for Ginkgo, and 16.9 years for Sand Hollow. Simple flow fields can be traced from vent to distal areas and an emplacement sequence visualized, but those with multiple-layered lobes present a degree of complexity that make lava pathways and emplacement sequences more difficult to identify.
Nuclear waste management and the impact of Carter Administration policies

International Nuclear Information System (INIS)

Williams, R.F.

1979-01-01

The impact of Carter Administration's policies on the nuclear waste management program are evaluated in this article. The waste management program faces numerous inconsistencies resulting from: a lack of a clearly defined schedule and division of responsibility; the requirement to meet conflicting procedural requirements; and the lack of clear statements from the President and Congress supporting the major programs. Some of the ramifications of these points are discussed with reference to the schedule for the 3 key program elements: National Waste Terminal Storage (NWTS) Facility scheduled for commercial operation in 1985 to handle commercial high-level wastes; the Spent Unreprocessed Fuel (SURF) Facility scheduled for operation in 1985 to handle spent fuel from commercial power plants; the Waste Isolation Pilot Plant (WIPP) scheduled for operation in 1983 or 1984 for the disposal of TRU defense wastes. Possible avenues for improvement are suggested
Logistics characterization for regional spent fuel repositories concept

International Nuclear Information System (INIS)

Joy, D.S.; Hudson, B.J.; Anthony, M.W.

1980-08-01

This report summarizes a study of logistics considerations for a four-region repository system for spent fuel disposal. The logistics considerations include: (1) yearly receipt and emplacement; (2) inventory; (3) away-from-reactor (AFR) storage; (4) nuclear capacity growth effects; (5) entire lifetime of reactors served by repository operations; (6) proportions of pressurized-water-reactor (PWR)/boiling-water-reactor (BWR) fuel; (7) proportions of rail and truck shipments; (8) shipping cask fleet requirements; (9) number of annual shipments; (10) mode (rail/truck) and cost of shipment; and (11) initial year for shipment to maintain full core reserve. The nation was divided into Northeast, North Central, Southern, and Western regions for evaluation purposes. Repository logistics were analyzed in each region based on three different capacity projections. For the Southern region, results for seven salt dome sites are presented. The Western region results cover four potential sites. The North Central and Northeastern regions results are not presented on a site specific basis. Conclusions are drawn based on the results. The methodology assumptions and references used in the logistics analysis are described for the convenience of the reader
Initial results from dissolution rate testing of N-Reactor spent fuel over a range of potential geologic repository aqueous conditions

International Nuclear Information System (INIS)

Gray, W.J.; Einziger, R.E.

1998-04-01

Hanford N-Reactor spent nuclear fuel (HSNF) may ultimately be placed in a geologic repository for permanent disposal. To determine whether the engineered barrier system that will be designed for emplacement of light-water-reactor (LWR) spent fuel will also suffice for HSNF, aqueous dissolution rate measurements were conducted on the HSNF. The purpose of these tests was to determine whether HSNF dissolves faster or slower than LWR spent fuel under some limited repository-relevant water chemistry conditions. The tests were conducted using a flowthrough method that allows the dissolution rate of the uranium matrix to be measured without interference by secondary precipitation reactions that would confuse interpretation of the results. Similar tests had been conducted earlier with LWR spent fuel, thereby allowing direct comparisons. Two distinct corrosion modes were observed during the course of these 12 tests. The first, Stage 1, involved no visible corrosion of the test specimen and produced no undissolved corrosion products. The second, Stage 2, resulted in both visible corrosion of the test specimen and left behind undissolved corrosion products. During Stage 1, the rate of dissolution could be readily determined because the dissolved uranium and associated fission products remained in solution where they could be quantitatively analyzed. The measured rates were much faster than has been observed for LWR spent fuel under all conditions tested to date when normalized to the exposed test specimen surface areas. Application of these results to repository conditions, however, requires some comparison of the physical conditions of the different fuels. The surface area of LWR fuel that could potentially be exposed to repository groundwater is estimated to be approximately 100 times greater than HSNF. Therefore, when compared on the basis of mass, which is more relevant to repository conditions, the HSNF and LWR spent fuel dissolve at similar rates
Containing method for spent fuel and spent fuel containing vessel

International Nuclear Information System (INIS)

Maekawa, Hiromichi; Hanada, Yoshine.

1996-01-01

Upon containing spent fuels, a metal vessel main body and a support spacer having fuel containing holes are provided. The support spacer is disposed in the inside of the metal vessel main body, and spent fuel assemblies are loaded in the fuel containing holes. Then, a lid is welded at the opening of the metal vessel main body to provide a sealing state. In this state, heat released from the spent fuel assemblies is transferred to the wall of the metal vessel main body via the support spacer. Since the support spacer has a greater heat conductivity than gases, heat of the spent fuel assemblies tends to be released to the outside, thereby capable of removing heat of the spent fuel assemblies effectively. In addition, since the surfaces of the spent fuel assemblies are in contact with the inner surface of the fuel containing holes of the support spacer, impact-resistance and earthquake-resistance are ensured, and radiation from the spent fuel assemblies is decayed by passing through the layer of the support spacer. (T.M.)
Wireless Sensor Networks for Detection of IED Emplacement

Science.gov (United States)

2009-06-01

unclassified Standard Form 298 (Rev. 8-98) Prescribed by ANSI Std Z39-18 Abstract We are investigating the use of wireless nonimaging -sensor...networks for the difficult problem of detection of suspicious behavior related to IED emplacement. Hardware for surveillance by nonimaging -sensor networks...with people crossing a live sensor network. We conclude that nonimaging -sensor networks can detect a variety of suspicious behavior, but
Statistical analysis of the sustained lava dome emplacement and destruction processes at Popocatépetl volcano, Central México

Science.gov (United States)

Mendoza-Rosas, Ana Teresa; Gómez-Vázquez, Ángel; De la Cruz-Reyna, Servando

2017-06-01

Popocatépetl volcano reawakened in 1994 after nearly 70 years of quiescence. Between 1996 and 2015, a succession of at least 38 lava domes have been irregularly emplaced and destroyed, with each dome reaching particular volumes at specific emplacement rates. The complexity of this sequence is analyzed using statistical methods in an attempt to gain insight into the physics and dynamics of the lava dome emplacement and destruction process and to objectively assess the hazards related to that volcano. The time series of emplacements, dome residences, lava effusion lulls, and emplaced dome volumes and thicknesses are modeled using the simple exponential and Weibull distributions, the compound non-homogeneous generalized Pareto-Poisson process (NHPPP), and the mixture of exponentials distribution (MOED). The statistical analysis reveals that the sequence of dome emplacements is a non-stationary, self-regulating process most likely controlled by the balance between buoyancy-driven magma ascent and volatile exsolution crystallization. This balance has supported the sustained effusive activity for decades and may persist for an undetermined amount of time. However, the eruptive history of Popocatépetl includes major Plinian phases that may have resulted from a breach in that balance. Certain criteria to recognize such breaching conditions are inferred from this statistical analysis.

Buffered and unbuffered dike emplacement on Earth and Venus - Implications for magma reservoir size, depth, and rate of magma replenishment

Science.gov (United States)

Parfitt, E. A.; Head, J. W., III

1993-01-01

Models of the emplacement of lateral dikes from magma chambers under constant (buffered) driving pressure conditions and declining (unbuffered) driving pressure conditions indicate that the two pressure scenarios lead to distinctly different styles of dike emplacement. In the unbuffered case, the lengths and widths of laterally emplaced dikes will be severely limited and the dike lengths will be highly dependent on chamber size; this dependence suggests that average dike length can be used to infer the dimensions of the source magma reservoir. On Earth, the characteristics of many mafic-dike swarms suggest that they were emplaced in buffered conditions (e.g., the Mackenzie dike swarm in Canada and some dikes within the Scottish Tertiary). On Venus, the distinctive radial fractures and graben surrounding circular to oval features and edifices on many size scales and extending for hundreds to over a thousand km are candidates for dike emplacement in buffered conditions.
STRUCTURAL CALCULATION OF AN EMPLACEMENT PALLET STATICALLY LOADED BY A WASTE PACKAGE

International Nuclear Information System (INIS)

S. Mastilovic

2000-01-01

The purpose of this calculation is to determine the structural response of the emplacement pallet (EP) subjected to static load from the mounted waste package (WP). The scope of this document is limited to reporting the calculation results in terms of stress intensity magnitudes. This calculation is associated with the waste emplacement systems design; calculations are performed by the Waste Package Design group. AP-3.12Q, Revision 0, ICN 0, Calculations, is used to perform the calculation and develop the document. The finite element solutions are performed by using the commercially available ANSYS Version (V) 5.4 finite element code. The results of these calculations are provided in terms of maximum stress intensity magnitudes
Study on a transportation and emplacement system of pre-assembled EBS module for HLW geological disposal

International Nuclear Information System (INIS)

Awano, Toshihiko; Kanno, Takeshi; Katsumata, Syunsuke; Kosuge, Kazuhiro

2009-01-01

HLW disposal is one of the largest issue to utilize Nuclear power safely. In the past study, the concept, which buffer materials and Overpacked waste were transported into underground respectively, have shown. The concept of pre-assembled engineered barrier has advantage to simplify the logistics and emplacement procedure, however there are difficulties to support heavy weight of pre-assembled package by equipment under the condition of little clearance between tunnel and package. In this study, Combination of air bearing and two degree-of-freedom wheels were suggested for transportation, and air jack was suggested for unloading and emplacement system. Also, whole system for transportation and emplacement procedure was designed, and Scale model test was examined to evaluate the feasibility of these concept and functions. (author)
Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

Energy Technology Data Exchange (ETDEWEB)

Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

2008-12-15

The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments
Quarter-scale modeling of room convergence effects on CH [contact-handled] TRU drum waste emplacements using WIPP [Waste Isolation Pilot Plant] reference design geometries

International Nuclear Information System (INIS)

VandeKraats, J.

1987-11-01

This study investigates the effect of horizontal room convergence on CH waste packages emplaced in the WIPP Reference Design geometry (rooms 13 feet high by 33 feet wide, with minus 3/8 inch screened backfill emplaced over and around the waste packages) as a function of time. Based on two tests, predictions were made with regard to full-scale 6-packs emplaced in the Reference Design geometry. These are that load will be transmitted completely through the stack within the first five years after waste emplacement and all drums in all 6-packs will be affected; that virtually all drums will show some deformation eight years after emplacement; that some drums may breach before the eighth year after emplacement has elapsed; and that based on criteria developed during testing, it is predicted that 1% of the drums emplaced will be breached after 8 years and, after 15 years, approximately 12% of the drums are predicted to be breached. 8 refs., 41 figs., 3 tabs
Performance implications of waste package emplacement orientation

International Nuclear Information System (INIS)

Wilder, D.G.

1991-05-01

Emplacement borehole orientation directly impacts many aspects of the Engineered Barrier System (EBS) and interactions with the near field environment. This paper considers the impacts of orientation on the hydrologic portion of the environment and its interactions with the EBS. The hydrologic environment is considered from a conceptual standpoint, the numerical analyses are left for subsequent work. As reported in this paper, several aspects of the hydrological environment are more favorable for long term performance of vertically oriented rather than horizontally oriented Waste Packages. 19 refs., 15 figs
National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Special study No. 1. Twenty-five-year retrievability, decommissioning cost estimate

International Nuclear Information System (INIS)

1978-11-01

This estimate covers decommissioning costs of facilities for the receipt, handling, and storage in bedded salt of canistered spent fuel assemblies from both BWR and PWR commercial power plants. The estimate includes all decommissioning costs for a repository, based on 25-year retrievability, constructed in accordance with the design shown in Conceptual Design Report (CDR), as modified by Special Study No. 1 (KE Report No. 78-60-RE) and decommissioned in accordance with the program outlined in the Conceptual Design Description Report, KE Report No. 78-58-R. Costs for Operating Contractor personnel on the site at this time are included in this report and not in the Operating Cost Estimate (KE Report 78-63-RE). The operating cost estimates end with the completion of storage room backfilling. The three major elements of decommissioning are: demolition of surface facilities, backfilling of main entries and airways, and shaft liner removal and shaft plugging. EDIT, ECON, and DELOX computer programs and a chart of accounts were furnished by UCC-ND under direction of the Government and the line item-capital-cost estimate was prepared according to the prescribed format. The decommissioning cost estimate referenced herein is in the same format as its companion line item-capital-cost estimate KE Report 78-62-RE
Spent Fuel Test-Climax: technical measurements data management system description and data presentation

International Nuclear Information System (INIS)

Carlson, R.C.

1985-08-01

The Spent Fuel Test-Climax (SFT-C) was located 420 m below surface in the Climax Stock granite on the Nevada Test Site. The test was conducted under the technical direction of the Lawrence Livermore National Laboratory (LLNL) as part of the Nevada Nuclear Waste Storage Investigations (NNWSI) for the US Department of Energy. Eleven canisters of spent nuclear reactor fuel were emplaced, along with six electrical simulators, in April-May 1980. The spent fuel canisters were retrieved and the electrical simulators de-energized in March-April 1983. During the test, just over 1000 MW-hr of thermal energy was deposited in the site, causing temperature changes 100 0 C near the canisters, and about 5 0 in the tunnels. More than 900 channels of geotechnical, seismological, and test status data were recorded on nearly continuous basis for about 3-1/2 years, ending in September 1983. Most geotechnical instrumentation was known to be temperature sensitive, and thus would require temperature compensation before interpretation. Accordingly, a 10-in. reel of digital tape was off-loaded and shipped to Livermore every 4 to 8 weeks, where the data were verified, organized into 45 one-million-word files, and temperature corrected. The purpose of this report is to document the receipt and processing of the data by LLNL Livermore personnel, present facts about the history of the instruments which may be important to the interpretation of the data, present the data themselves in graphical form for each instrument over its operating lifetime, document the forms and locations in which the data will be archived, and offer the data to the geotechnical community for future use in understanding and predicting the effects of the storage of heat-generating waste in hard rocks such as granite
Evolution of repository and waste package designs for Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste

International Nuclear Information System (INIS)

Rechard, Rob P.; Voegele, Michael D.

2014-01-01

This paper summarizes the evolution of the engineered barrier design for the proposed Yucca Mountain disposal system. Initially, the underground facility used a fairly standard panel and drift layout excavated mostly by drilling and blasting. By 1993, the layout of the underground facility was changed to accommodate construction by a tunnel boring machine. Placement of the repository in unsaturated zone permitted an extended period without backfilling; placement of the waste package in an open drift permitted use of much larger, and thus hotter packages. Hence in 1994, the underground facility design switched from floor emplacement of waste in small, single walled stainless steel or nickel alloy containers to in-drift emplacement of waste in large, double-walled containers. By 2000, the outer layer was a high nickel alloy for corrosion resistance and the inner layer was stainless steel for structural strength. Use of large packages facilitated receipt and disposal of high volumes of spent nuclear fuel. In addition, in-drift package placement saved excavation costs. Options considered for in-drift emplacement included different heat loads and use of backfill. To avoid dripping on the package during the thermal period and the possibility of localized corrosion, titanium drip shields were added for the disposal drifts by 2000. In addition, a handling canister, sealed at the reactor to eliminate further handling of bare fuel assemblies, was evaluated and eventually adopted in 2006. Finally, staged development of the underground layout was adopted to more readily adjust to changes in waste forms and Congressional funding. - Highlights: • Progression of events associated with repository design to accommodate tunnel boring machine and in-drift waste package emplacement are discussed. • Change in container design from small, single-layered stainless steel vessel to large, two-layered nickel alloy vessel is discussed. • The addition of drip shield to limit the
Analogue experiments as benchmarks for models of lava flow emplacement

Science.gov (United States)

Garel, F.; Kaminski, E. C.; Tait, S.; Limare, A.

2013-12-01

During an effusive volcanic eruption, the crisis management is mainly based on the prediction of lava flow advance and its velocity. The spreading of a lava flow, seen as a gravity current, depends on its "effective rheology" and on the effusion rate. Fast-computing models have arisen in the past decade in order to predict in near real time lava flow path and rate of advance. This type of model, crucial to mitigate volcanic hazards and organize potential evacuation, has been mainly compared a posteriori to real cases of emplaced lava flows. The input parameters of such simulations applied to natural eruptions, especially effusion rate and topography, are often not known precisely, and are difficult to evaluate after the eruption. It is therefore not straightforward to identify the causes of discrepancies between model outputs and observed lava emplacement, whereas the comparison of models with controlled laboratory experiments appears easier. The challenge for numerical simulations of lava flow emplacement is to model the simultaneous advance and thermal structure of viscous lava flows. To provide original constraints later to be used in benchmark numerical simulations, we have performed lab-scale experiments investigating the cooling of isoviscous gravity currents. The simplest experimental set-up is as follows: silicone oil, whose viscosity, around 5 Pa.s, varies less than a factor of 2 in the temperature range studied, is injected from a point source onto a horizontal plate and spreads axisymmetrically. The oil is injected hot, and progressively cools down to ambient temperature away from the source. Once the flow is developed, it presents a stationary radial thermal structure whose characteristics depend on the input flow rate. In addition to the experimental observations, we have developed in Garel et al., JGR, 2012 a theoretical model confirming the relationship between supply rate, flow advance and stationary surface thermal structure. We also provide
Effect of repository underground ventilation on emplacement drift temperature control

International Nuclear Information System (INIS)

Yang, H.; Sun, Y.; McKenzie, D.G.; Bhattacharyya, K.K.

1996-01-01

The repository advanced conceptual design (ACD) is being conducted by the Civilian Radioactive Waste Management System, Management ampersand Operating Contractor. Underground ventilation analyses during ACD have resulted in preliminary ventilation concepts and design methodologies. This paper discusses one of the recent evaluations -- effects of ventilation on emplacement drift temperature management
Spent fuel characterization for the commercial waste and spent fuel packaging program

International Nuclear Information System (INIS)

Fish, R.L.; Davis, R.B.; Pasupathi, V.; Klingensmith, R.W.

1980-03-01

This document presents the rationale for spent fuel characterization and provides a detailed description of the characterization examinations. Pretest characterization examinations provide quantitative and qualitative descriptions of spent fuel assemblies and rods in their irradiated conditions prior to disposal testing. This information is essential in evaluating any subsequent changes that occur during disposal demonstration and laboratory tests. Interim examinations and post-test characterization will be used to identify fuel rod degradation mechanisms and quantify degradation kinetics. The nature and behavior of the spent fuel degradation will be defined in terms of mathematical rate equations from these and laboratory tests and incorporated into a spent fuel performance prediction model. Thus, spent fuel characterization is an essential activity in the development of a performance model to be used in evaluating the ability of spent fuel to meet specific waste acceptance criteria and in evaluating incentives for modification of the spent fuel assemblies for long-term disposal purposes
Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

International Nuclear Information System (INIS)

Johnson, G.L.

1991-11-01

Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. On such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97 degrees C and whether the cladding of the stored spent fuel ever exceeds 350 degrees C. Limiting the borehole to temperatures of 97 degrees C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350 degrees C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97 degrees C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350 degrees C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft x 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40 degrees C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation
Salt Repository emplacement mode evaluation and selection: Final report

International Nuclear Information System (INIS)

1988-03-01

This document describes the decision analysis performed to evaluate and compare the emplacement mode for the Salt Repository. The study was commissioned to recommend one emplacement mode to the Salt Repository Project Office using multi-attribute decision analysis. The nature of the decision required analysis of uncertain outcomes and conflicting attributes and offers a high degree of objectivity for these types of decisions since the decision model is structured to allow only the facts to enter into the final decision. The analysis requires an explicit definition of the attributes used to evaluate the alternative (e.g., cost, safety, environmental impact), the definition of a utility function over the attributes which incorporated both risk attitudes and trade-offs between attributes, and the probability distribution over the outcomes that would result from the selection of one alternative over the other. The decision process is described and results are given. A simulation model was developed to evaluate the probability distributions over the attributes. This report documents logic, inputs and results of this model. Final ranking of alternatives is given. Extensive technical backup documentation is included in the appendices to provide the quantitative basis for this decision. 5 refs., 2 figs., 8 tabs
Safety assessment document for spent fuel handling, packaging, and storage demonstrations at the E-MAD facility on the Nevada Test Site

International Nuclear Information System (INIS)

1985-04-01

The objectives for spent fuel handling and packaging demonstration are to develop the capability to satisfactorily encapsulate typical commercial nuclear reactor spent fuel assemblies and to establish the suitability of interim dry surface and near surface storage concepts. To accomplish these objectives, spent fuel assemblies from a pressurized water reactor have been received, encapsulated in steel canisters, and emplaced in on-site storage facilities and subjected to other tests. As an essential element of these demonstrations, a thorough safety assessment of the demonstration activities conducted at the E-MAD facility has been completed. This document describes the site location and characteristics, the existing E-MAD facility, and the facility modifications and equipment additions made specifically for the demonstrations. The document also summarizes the Quality Assurance Program utilized, and specifies the principal design criteria applicable to the facility modifications, equipment additions, and process operations. Evaluations have been made of the radiological impacts of normal operations, abnormal operations, and postulated accidents. Analyses have been performed to determine the affects on nuclear criticality safety of postulated accidents and credible natural phenomena. The consequences of postulated accidents resulting in fission product gas release have also been estimated. This document identifies the engineered safety features, procedures, and site characteristics that (1) prevent the occurrence of potential accidents or (2) assure that the consequences of postulated accidents are either insignificant or adequately mitigated
Preliminary conceptual designs for advanced packages for the geologic disposal of spent fuel

Energy Technology Data Exchange (ETDEWEB)

Westerman, R.E.

1979-04-01

The present study assumes that the spent fuel will be disposed of in mined repositories in continental geologic formations, and that the post-emplacement control of the radioactive species will be accomplished independently by both the natural barrier, i.e., the geosphere, and the engineered barrier system, i.e., the package components consisting of the stabilizer, the canister, and the overpack; and the barrier components external to the package consisting of the hole sleeve and the backfill medium. The present document provides an overview of the nature of the spent fuel waste; the general approach to waste containment, using the defense-in-depth philosophy; material options, both metallic and nonmetallic, for the components of the engineered barrier system; a set of strawman criteria to guide the development of package/engineered barrier systems; and four preliminary concepts representing differing approaches to the solution of the containment problem. These concepts use: a corrosion-resistant meta canister in a special backfill (2 barriers); a mild steel canister in a corrosion-resistant metallic or nonmetallic hole sleeve, surrounded by a special backfill (2 barriers); a corrosion-resistant canister and a corrosion-resistant overpack (or hole sleeve) in a special backfill (3 barriers); and a mild steel canister in a massive corrosion-resistant bore sleeve surrounded by a polymer layer and a special backfill (3 barriers). The lack of definitive performance requirements makes it impossible to evaluate these concepts on a functional basis at the present time.
Safety Consideration for a Wet Interim Spent Fuel Store at Conceptual Design Stage

International Nuclear Information System (INIS)

Astoux, Marion

2014-01-01

EDF Energy plans to build and operate two UK EPRs at the Hinkley Point C (HPC) site in Somerset, England. Spent fuel from the UK EPRs will need to be managed from the time it is discharged from the reactor until it is ultimately disposed of and this will involve storing the spent fuel for a period in the fuel building and thereafter in a dedicated interim facility until it can be emplaced within the UK Geological Disposal Facility. EDF Energy has proposed that this interim store should be located on the Hinkley Point site which is consistent with UK policy. This Interim Spent Fuel Store (ISFS) will have the capability to store for at least one hundred years the spent fuel arising from the operation of the two EPR units (sixty years operation). Therefore, specificities regarding the lifetime of the facility have to be accounted for its design. The choice of interim storage technology was considered in some depth for the HPC project and wet storage (pool) was selected. The facility is currently at conceptual design stage, although its construction will be part of main site construction phase. Safety functions and safety requirements for this storage facility have been defined, in compliance with WENRA 'Waste and Spent Fuel Storage - Safety Reference Level Report' and IAEA Specific Safety Guide no. 15 'Storage of Spent Nuclear Fuel'. EDF technical know-how, operational feedback on existing storage pools, UK regulatory context and Fukushima experience feedback have also been accounted for. Achievement of the safety functions as passively as reasonably practicable is a key issue for the design, especially in accident situations. Regarding lifetime aspects, ageing management of equipments, optimisation of the refurbishment, climate change, passivity of the facility, and long-term achievement of the safety functions are among the subjects to consider. Adequate Operational Limits and Conditions will also have to be defined, to enable the long-term achievement of the safety
Spent fuel management

International Nuclear Information System (INIS)

2005-01-01

The production of nuclear electricity results in the generation of spent fuel that requires safe, secure and efficient management. Appropriate management of the resulting spent fuel is a key issue for the steady and sustainable growth of nuclear energy. Currently about 10,000 tonnes heavy metal (HM) of spent fuel are unloaded every year from nuclear power reactors worldwide, of which 8,500 t HM need to be stored (after accounting for reprocessed fuel). This is the largest continuous source of civil radioactive material generated, and needs to be managed appropriately. Member States have referred to storage periods of 100 years and even beyond, and as storage quantities and durations extend, new challenges arise in the institutional as well as in the technical area. The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs
Waste package/engineered barrier system design concepts for the direct disposal of spent fuel in the potential United States' repository at Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Stahl, D.; Harrison, D.J.

1993-01-01

The goal of the US Department of Energy's (DOE) Yucca Mountain Site Characterization Project (YMP) waste package development program is to design a waste package and associated engineered barrier system (EBS) that meets the applicable regulatory requirements for safe disposal of spent nuclear fuel and solidified high-level waste (HLW) in a geologic repository. Attainment of this goal relies on a multi-barrier approach, the unsaturated nature of the Yucca Mountain site, consideration of technical alternatives, and sufficient resolution of technical and regulatory uncertainties. To accomplish this, an iterative system engineering approach will be used. The NWPA of 1982 limits the content of the first US repository to 70,000 metric tons of heavy metal (MTHM). The DOE Mission Plan describes the implementation of the provisions of the NWPA for the waste management system. The Draft 1988 approach will involve selecting candidate designs, evaluating them against performance requirements, and then selecting one or two preferred designs for further detailed evaluation and final design. The reference design of the waste package described in the YMP Site Characterization Plan is a thin-walled, vertical borehole-emplaced waste package with an air gap between the package and the rock wall. The reference design appeared to meet the design requirement. However, the degree of uncertainty was large. This uncertainty led to considering several more-robust design concepts during the Advanced Conceptual Design phase of the program that include small, drift-emplaced packages and higher capacity, drift-emplaced packages, both partially and totally self-shielded. Metallic as well as ceramic materials are being considered
Spent fuel workshop'2002

International Nuclear Information System (INIS)

Poinssot, Ch.

2002-01-01

This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO 2 fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO 2 dissolution determined from electrochemical experiments with 238 Pu doped UO 2 M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO 2 studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with α doped UO 2 in Boom clay conditions (K. Lemmens), Studies of the behavior of UO 2 / water interfaces under He 2+ beam (C. Corbel), Alpha and gamma radiolysis effects on UO 2 alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines (M. Kelm), On the potential catalytic behavior of

Shapes of Venusian 'pancake' domes imply episodic emplacement and silicic composition

Science.gov (United States)

Fink, Jonathan H.; Bridges, Nathan T.; Grimm, Robert E.

1993-01-01

The main evidence available for constraining the composition of the large circular 'pancake' domes on Venus is their gross morphology. Laboratory simulations using polyethylene glycol show that the height to diameter (aspect) ratios of domes of a given total volume depend critically on whether their extrusion was continuous or episodic, with more episodes leading to greater cooling and taller domes. Thus without observations of their emplacement, the compositions of Venusian domes cannot be uniquely constrained by their morphology. However, by considering a population of 51 Venusian domes to represent a sampling of many stages during the growth of domes with comparable histories, and by plotting aspect ratio versus total volume, we find that the shapes of the domes are most consistent with episodic emplacement. On Earth this mode of dome growth is found almost exclusively in lavas of dacite to rhyolite composition, strengthening earlier inferences about the presence of evolved magmas on Venus.
Additive Construction with Mobile Emplacement (ACME)

Science.gov (United States)

Vickers, John

2015-01-01

The Additive Construction with Mobile Emplacement (ACME) project is developing technology to build structures on planetary surfaces using in-situ resources. The project focuses on the construction of both 2D (landing pads, roads, and structure foundations) and 3D (habitats, garages, radiation shelters, and other structures) infrastructure needs for planetary surface missions. The ACME project seeks to raise the Technology Readiness Level (TRL) of two components needed for planetary surface habitation and exploration: 3D additive construction (e.g., contour crafting), and excavation and handling technologies (to effectively and continuously produce in-situ feedstock). Additionally, the ACME project supports the research and development of new materials for planetary surface construction, with the goal of reducing the amount of material to be launched from Earth.
Thermal, operational, and economic aspects of repository design alternatives

International Nuclear Information System (INIS)

Closs, K.D.; Papp, R.; Bechthold, W.; Engelmann, H.J.; Hartje, B.

1989-01-01

This paper discusses how a broad spectrum of heat-generating nuclear waste types like reprocessing waste (HLW, ILW), spent LWR, and spent HTR fuel will be disposed of in a future German repository located in a salt dome. Different package and emplacement concepts for the various waste forms are feasible, ranging from pure borehole emplacement of canisters and drums to pure drift emplacement of heavily shielded casks as well as combinations of both concepts. Optimization of the whole back-end system (waste treatment, interim storage, and disposal) is performed taking into account thermal, operational, and economic aspects. From a radiological viewpoint, the drift emplacement concept is superior to all concepts in as much as the occupational dose is lowest and zero release from the casks is attainable during normal and anomalous repository operation. As far as economics are concerned, the cost of the heavily shielded casks necessary for drift emplacement contributes markedly to the overall cost of the whole back-end system. Drift emplacement can compete economically with the other concepts only if the casks are used both for interim storage and disposal
High level radioactive waste repositories. Task 3. Review of underground handling and emplacement. 1. Executive summary

Energy Technology Data Exchange (ETDEWEB)

1981-01-01

A review is presented of proposals for transport, handling and emplacement of high-level radioactive waste in an underground repository appropriate to the U.K. context, with particular reference to waste block size and configuration; self-shielded or partially-shielded block; stages of disposal; transport by road/rail to repository site; handling techniques within repository; emplacement in vertical holes or horizontal tunnels; repository access by adit, incline or shaft; conventional and radiological safety; costs; and major areas of uncertainty requiring research or development.
Vertical boreholes in salt - An industrial demonstration program -

International Nuclear Information System (INIS)

Bollingerfehr, W.; Filbert, W.

2006-01-01

In order to synchronize and optimize the emplacement technologies used for both categories of waste (vitrified waste and spent fuel) the borehole emplacement technique for consolidated spent fuel was reconsidered in Germany. The appropriate design resulted in a fuel rod canister of the type 'BSK 3'. This BSK 3 steel canister is of the same diameter as the standard HLW-canister which allows the use of a common transfer and handling technique. The canister is tightly sealed by welding and designed to withstand petrostatic pressure. Thermal calculations showed that the emplacement of a BSK 3-canister into a vertical borehole in a salt repository is possible after only about 3 to 7 years after reactor unloading of the fuel assemblies. Thus, the emplacement of BSK 3-canisters allows a complete revision of the schedule of spent fuel disposal concepts and may lead to a considerable reduction of time and costs. Accordingly, a research program was launched in 2004 within the framework of the 6th European Research Program to develop the transport and emplacement components, the functionality and reliability of which are to be tested in a one-year demonstration phase which will commence at the beginning of 2008. (author)
From Embodiment to Emplacement: Re-Thinking Competing Bodies, Senses and Spatialities

Science.gov (United States)

Pink, Sarah

2011-01-01

In this article I discuss how a shift from theories of embodiment to one of emplacement can inform how we understand the performing body in competitive and pedagogical contexts. I argue that recent theoretical advances concerning the senses, human perception and place offer new analytical possibilities for understanding skilled performances and…
Emplacement time of Salai Patai carbonatite, Malakand, Pakistan, from fission track dating of zircon and apatite

International Nuclear Information System (INIS)

Qureshi, A.A.; Khan, H.A.

1991-01-01

Based on fission track dating of zircon and apatite, the emplacement history of Salai Patai carbonatite has been traced. It has been estimated that the carbonatite was emplaced along the thrust plane associated with the Indian-Eurasian plate collision during the Oligocene period followed by some thermal/tectonic episode during Early Miocene. This negates the previous proposal that all carbonatites found in Pakistan are a part of a 200 km long alkaline province associated with the rifting of Peshawar Valley during Late Cretaceous or early tertiary. (author)
As-built mechanical and thermomechanical calculations of a spent-fuel test in Climax Stock granite

International Nuclear Information System (INIS)

Butkovich, T.R.

1981-09-01

A generic test of geological storage of spent-fuel assemblies from an operating nuclear reactor in Climax Stock granite has been underway at the US Department of Energy's Nevada Test Site since spring 1980. The spent-fuel assemblies were emplaced in the floor of the central drift of three parallel drifts. Auxiliary electric heaters were installed in the floors of the side drifts to simulate a large repository. Calculational modeling of the spent-fuel repository was made with the finite element codes, ADINA stress analysis code and the compatible ADINAT heat flow code. The primary purpose for doing the calculations was to provide results with the best estimates of physical and mechanical rock properties and in situ stresses. Field measurements show the effective modulus of Climax Stock granite is a factor of two lower than that obtained in the laboratory on small samples. Comparative calculations using these field measurements and measurements obtained in the laboratory show displacements are approximately inversely proportional to the modulus. They also indicate the importance of knowing the effective in situ modulus. Another comparison, varying the in situ stress loading, shows the importance of knowing the correct value of in situ stress. Calculations using the best-estimate values of rock properties, effective in situ modulus and Poisson's ratio, and in situ stress are not intended to predict the displacement and stress changes from the mining and the imposition of a thermal load. The existing jointing in the Climax Stock granite could have a significant effect on the actual displacements and stress field. Such effects are not accounted for here. However, these calculations do indicate the direction and magnitude of displacements and stresses that would be expected in absence of jointing. Comparison of those results with actual field measurements will, thus, indicate the significance of including a joint model in future calculations
Guidebook on spent fuel storage

International Nuclear Information System (INIS)

1984-01-01

The Guidebook summarizes the experience and information in various areas related to spent fuel storage: technological aspects, the transport of spent fuel, economical, regulatory and institutional aspects, international safeguards, evaluation criteria for the selection of a specific spent fuel storage concept, international cooperation on spent fuel storage. The last part of the Guidebook presents specific problems on the spent fuel storage in the United Kingdom, Sweden, USSR, USA, Federal Republic of Germany and Switzerland
Spent fuels program

International Nuclear Information System (INIS)

Shappert, L.B.

1983-01-01

The goal of this task is to support the Domestic Spent Fuel Storage Program through studies involving the transport of spent fuel. A catalog was developed to provide authoritative, timely, and accessible transportation information for persons involved in the transport of irradiated reactor fuel. The catalog, drafted and submitted to the Transportation Technology Center, Sandia National Laboratories, for their review and approval, covers such topics as federal, state, and local regulations, spent fuel characteristics, cask characteristics, transportation costs, and emergency response information
Emplacement of rock avalanche material across saturated sediments, Southern Alp, New Zealand

Science.gov (United States)

Dufresne, A.; Davies, T. R.; McSaveney, M. J.

2012-04-01

The spreading of material from slope failure events is not only influenced by the volume and nature of the source material and the local topography, but also by the materials encountered in the runout path. In this study, evidence of complex interactions between rock avalanche and sedimentary runout path material were investigated at the 45 x 106 m3 long-runout (L: 4.8 km) Round Top rock avalanche deposit, New Zealand. It was sourced within myolinitic schists of the active strike-slip Alpine Fault. The narrow and in-failure-direction elongate source scarp is deep-seated, indicating slope failure was triggered by strong seismic activity. The most striking morphological deposit features are longitudinal ridges aligned radially to source. Trenching and geophysical surveys show bulldozed and sheared substrate material at ridge termini and laterally displaced sedimentary strata. The substrate failed at a minimum depth of 3 m indicating a ploughing motion of the ridges into the saturated material below. Internal avalanche compression features suggest deceleration behind the bulldozed substrate obstacle. Contorted fabric in material ahead of the ridge document substrate disruption by the overriding avalanche material deposited as the next down-motion hummock. Comparison with rock avalanches of similar volume but different emplacement environments places Round Top between longer runout avalanches emplaced over e.g. playa lake sediments and those with shorter travel distances, whose runout was apparently retarded by topographic obstacles or that entrained high-friction debris. These empirical observations indicate the importance of runout path materials on tentative trends in rock avalanche emplacement dynamics and runout behaviour.
Subsurface geology of a potential waste emplacement site, Salt Valley Anticline, Grand County, Utah

Science.gov (United States)

Hite, R.J.

1977-01-01

The Salt Valley anticline, which is located about 32 km northeast of Moab, Utah, is perhaps one of the most favorable waste emplacement sites in the Paradox basin. The site, which includes about 7.8 km 2, is highly accessible and is adjacent to a railroad. The anticline is one of a series of northwest-trending salt anticlines lying along the northeast edge of the Paradox basin. These anticlines are cored by evaporites of the Paradox Member of the Hermosa Formation of Middle Pennsylvanian age. The central core of the Salt Valley anticline forms a ridgelike mass of evaporites that has an estimated amplitude of 3,600 m. The evaporite core consists of about 87 percent halite rock, which includes some potash deposits; the remainder is black shale, silty dolomite, and anhydrite. The latter three lithologies are referred to as 'marker beds.' Using geophysical logs from drill holes on the anticline, it is possible to demonstrate that the marker beds are complexly folded and faulted. Available data concerning the geothermal gradient and heatflow at the site indicate that heat from emplaced wastes should be rapidly dissipated. Potentially exploitable resources of potash and petroleum are present at Salt Valley. Development of these resources may conflict with use of the site for waste emplacement.
Subsurface geology of a potential waste emplacement site, Salt Valley Anticline, Grand County, Utah

International Nuclear Information System (INIS)

Hite, R.J.

1977-01-01

The Salt Valley anticline, which is located about 32 km northeast of Moab, Utah, is perhaps one of the most favorable waste emplacement sites in the Paradox basin. The site, which includes about 7.8 km 2 , is highly accessible and is adjacent to a railroad. The anticline is one of a series of northwest-trending salt antilcines lying along the northeast edge of the Paradox basin. These anticlines are cored by evaporites of the Paradox Member of the Hermosa Formation of Middle Pennsylvanian age. The central core of the Salt Valley anticline forms a ridgelike mass of evaporites that has an estimated amplitude of 3,600 m. The evaporite core consists of about 87 percent halite rock, which includes some potash deposits; the remainder is black shale, silty dolomite, and anhydrite. The latter three lithologies are referred to as ''marker beds.'' Using geophysical logs from drill holes on the anticline, it is possible to demonstrate that the marker beds are complexly folded and faulted. Available data concerning the geothermal gradient and heatflow at the site indicate that heat from emplaced wastes should be rapidly dissipated. Potentially exploitable resources of potash and petroleum are present at Salt Valley. Development of these resources may conflict with use of the site for waste emplacement
Final environmental statement: US Spent Fuel Policy. Storage of foreign spent power reactor fuel

International Nuclear Information System (INIS)

1980-05-01

In October 1977, the Department of Energy (DOE) announced a Spent Fuel Storage Policy for nuclear power reactors. Under this policy, as approved by the President, US utilities will be given the opportunity to deliver spent fuel to US Government custody in exchange for payment of a fee. The US Government will also be prepared to accept a limited amount of spent fuel from foreign sources when such action would contribute to meeting nonproliferation goals. Under the new policy, spent fuel transferred to the US Government will be delivered - at user expense - to a US Government-approved site. Foreign spent fuel would be stored in Interim Spent Fuel Storage (ISFS) facilities with domestic fuel. This volume of the environmental impact statement includes effects associated with implementing or not implementing the Spent Fuel Storage Policy for the foreign fuels. The analyses show that there are no substantial radiological health impacts whether the policy is implemented or not. In no case considered does the population dose commitment exceed 0.000006% of the world population dose commitment from natural radiation sources over the period analyzed. Full implementation of the US offer to accept a limited amount of foreign spent fuel for storage provides the greatest benefits for US nonproliferation policy. Acceptance of lesser quantities of foreign spent fuel in the US or less US support of foreign spent fuel storage abroad provides some nonproliferation benefits, but at a significantly lower level than full implementation of the offer. Not implementing the policy in regard to foreign spent fuel will be least productive in the context of US nonproliferation objectives. The remainder of the summary provides a brief description of the options that are evaluated, the facilities involved in these options, and the environmental impacts, including nonproliferation considerations, associated with each option
Spent nuclear fuel storage

International Nuclear Information System (INIS)

Romanato, Luiz Sergio

2005-01-01

When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)
Engineering for a disposal facility using the in-room emplacement method

Energy Technology Data Exchange (ETDEWEB)

Baumgartner, P; Bilinsky, D M; Ates, Y; Read, R S; Crosthwaite, J L; Dixon, D A

1996-06-01

This report describes three nuclear fuel waste disposal vaults using the in-room emplacement method. First, a generic disposal vault design is provided which is suitable for a depth range of 500 m to 1000 m in highly stressed, sparsely fractured rock. The design process is described for all components of the system. The generic design is then applied to two different disposal vaults, one at a depth of 750 m in a low hydraulically conductive, sparsely fractured rock mass and another at a depth of 500 m in a higher conductivity, moderately fractured rock mass. In the in-room emplacement method, the disposal containers with used-fuel bundles are emplaced within the confines of the excavated rooms of a disposal vault. The discussion of the disposal-facility design process begins with a detailed description of a copper-shell, packed-particulate disposal container and the factors that influenced its design. The disposal-room generic design is presented including the detailed specifications, the scoping and numerical thermal and thermal mechanical analyses, the backfilling and sealing materials, and the operational processes. One room design is provided that meets all the requirements for a vault depth range of 500 to 1000 m. A disposal-vault layout and the factors that influenced its design are also presented, including materials handling, general logistics, and separation of radiological and nonradiological operations. Modifications to the used-fuel packaging plant for the filling and sealing of the copper-shell, packed-particulate disposal containers and a brief description of the common surface facilities needed by the disposal vault and the packaging plant are provided. The implementation of the disposal facility is outlined, describing the project stages and activities and itemizing a specific plan for each of the project stages: siting, construction, operation; decommissioning; and closure. (author). 72 refs., 15 tabs., 63 figs.
Engineering for a disposal facility using the in-room emplacement method

International Nuclear Information System (INIS)

Baumgartner, P.; Bilinsky, D.M.; Ates, Y.; Read, R.S.; Crosthwaite, J.L.; Dixon, D.A.

1996-06-01

This report describes three nuclear fuel waste disposal vaults using the in-room emplacement method. First, a generic disposal vault design is provided which is suitable for a depth range of 500 m to 1000 m in highly stressed, sparsely fractured rock. The design process is described for all components of the system. The generic design is then applied to two different disposal vaults, one at a depth of 750 m in a low hydraulically conductive, sparsely fractured rock mass and another at a depth of 500 m in a higher conductivity, moderately fractured rock mass. In the in-room emplacement method, the disposal containers with used-fuel bundles are emplaced within the confines of the excavated rooms of a disposal vault. The discussion of the disposal-facility design process begins with a detailed description of a copper-shell, packed-particulate disposal container and the factors that influenced its design. The disposal-room generic design is presented including the detailed specifications, the scoping and numerical thermal and thermal mechanical analyses, the backfilling and sealing materials, and the operational processes. One room design is provided that meets all the requirements for a vault depth range of 500 to 1000 m. A disposal-vault layout and the factors that influenced its design are also presented, including materials handling, general logistics, and separation of radiological and nonradiological operations. Modifications to the used-fuel packaging plant for the filling and sealing of the copper-shell, packed-particulate disposal containers and a brief description of the common surface facilities needed by the disposal vault and the packaging plant are provided. The implementation of the disposal facility is outlined, describing the project stages and activities and itemizing a specific plan for each of the project stages: siting, construction, operation; decommissioning; and closure. (author)
Spent fuel storage facility, Kalpakkam

International Nuclear Information System (INIS)

Shreekumar, B.; Anthony, S.

2017-01-01

Spent Fuel Storage Facility (SFSF), Kalpakkam is designed to store spent fuel arising from PHWRs. Spent fuel is transported in AERB qualified/authorized shipping cask by NPCIL to SFSF by road or rail route. The spent fuel storage facility at Kalpakkam was hot commissioned in December 2006. All systems, structures and components (SSCs) related to safety are designed to meet the operational requirements
Spent fuel storage and isolation

International Nuclear Information System (INIS)

Bensky, M.S.; Kurzeka, W.J.; Bauer, A.A.; Carr, J.A.; Matthews, S.C.

1979-02-01

The principal spent fuel activities conducted within the commercial waste and spent fuel within the Commercial Waste and Spent Fuel Packaging Program are: simulated near-surface (drywell) storage demonstrations at Hanford and the Nevada Test Site; surface (sealed storage cask) and drywell demonstrations at the Nevada Test Site; and spent fuel receiving and packaging facility conceptual design. These investigations are described
Magnetic fabrics in characterization of magma emplacement and tectonic evolution of the Moyar Shear Zone, South India

Directory of Open Access Journals (Sweden)

P. Pratheesh

2013-01-01

Full Text Available The Moyar Shear Zone (MSZ of the South Indian granulite terrain hosts a prominent syenite pluton (∼560 Ma and associated NW-SE to NE-SW trending mafic dyke swarm (∼65 Ma and 95 Ma. Preliminary magnetic fabric studies in the mafic dykes, using Anisotropy of Magnetic Susceptibly (AMS studies at low-field, indicate successive emplacement and variable magma flow direction. Magnetic lineation and foliation in these dykes are identical to the mesoscopic fabrics in MSZ mylonites, indicating shear zone guided emplacement. Spatial distribution of magnetic lineation in the dykes suggests a common conduit from which the source magma has been migrated. The magnetic foliation trajectories have a sigmoidal shape to the north of the pluton and curve into the MSZ suggesting dextral sense of shear. Identical fabric conditions for magnetic fabrics in the syenite pluton and measured field fabrics in mylonite indicate syntectonic emplacement along the Proterozoic crustal scale dextral shear zone with repeated reactivation history.

TURVA-2012 safety case for licensing a spent fuel repository at Olkiluoto, Finland

International Nuclear Information System (INIS)

Vira, Juhani; Snellman, Margit

2014-01-01

In 2001, the Finnish Parliament endorsed a decision-in-principle (DiP) whereby the spent nuclear fuel produced by the operating nuclear reactors at Olkiluoto and Loviisa will be disposed of in a geological repository at Olkiluoto, on the south-western coast of Finland. Subsequently, additional DiPs were issued allowing the extension of the repository to accommodate spent nuclear fuel from additional reactors that are under construction or in planning at Olkiluoto, which means a total of 9 000 tU of spent nuclear fuel to be disposed of. In accordance with the decision of the Ministry of Trade and Industry (KTM) in 2003, Posiva submitted an application for a license to construct a disposal facility at Olkiluoto in 2012, consisting of an encapsulation facility and an underground deep geological repository. The application included a Preliminary Safety Analysis Report (PSAR) and a long-term safety case, TURVA-2012. Assuming a positive outcome of the current licensing review, the next step would be the Final Safety Analysis Report (FSAR) in support of an operational licence application around 2020. The disposal method is based on the same KBS-3 concept that the Swedish SKB has used as basis for their license application in 2010. Accordingly, the spent nuclear fuel will be encapsulated in water- and gas-tight copper canisters equipped with a load-bearing insert and emplaced in a deep geological repository constructed in the bedrock. The canisters will be surrounded by a swelling clay buffer material that isolates them from the bedrock. The deposition tunnels and the central tunnels and the other underground openings will be backfilled with materials of low permeability. The repository will be at a depth of about 400-450 m below ground. The primary role of the bedrock is to provide sufficiently stable conditions for the engineered barrier system and to make inadvertent human intrusion unlikely. In case of EBS failure, the bedrock shall also retain and retard the possible
Spent fuel transportation in the United States: commercial spent fuel shipments through December 1984

International Nuclear Information System (INIS)

1986-04-01

This report has been prepared to provide updated transportation information on light water reactor (LWR) spent fuel in the United States. Historical data are presented on the quantities of spent fuel shipped from individual reactors on an annual basis and their shipping destinations. Specifically, a tabulation is provided for each present-fuel shipment that lists utility and plant of origin, destination and number of spent-fuel assemblies shipped. For all annual shipping campaigns between 1980 and 1984, the actual numbers of spent-fuel shipments are defined. The shipments are tabulated by year, and the mode of shipment and the casks utilized in shipment are included. The data consist of the current spent-fuel inventories at each of the operating reactors as of December 31, 1984. This report presents historical data on all commercial spent-fuel transportation shipments have occurred in the United States through December 31, 1984
Hanford 100-N Area In Situ Apatite and Phosphate Emplacement by Groundwater and Jet Injection: Geochemical and Physical Core Analysis

Energy Technology Data Exchange (ETDEWEB)

Szecsody, James E.; Vermeul, Vincent R.; Fruchter, Jonathan S.; Williams, Mark D.; Rockhold, Mark L.; Qafoku, Nikolla; Phillips, Jerry L.

2010-07-01

The purpose of this study is to evaluate emplacement of phosphate into subsurface sediments in the Hanford Site 100-N Area by two different technologies: groundwater injection of a Ca-citrate-PO4 solution and water-jet injection of sodium phosphate and/or fish-bone apatite. In situ emplacement of phosphate and apatite adsorbs, then incorporates Sr-90 into the apatite structure by substitution for calcium. Overall, both technologies (groundwater injection of Ca-citrate-PO4) and water-jet injection of sodium phosphate/fish-bone apatite) delivered sufficient phosphate to subsur¬face sediments in the 100-N Area. Over years to decades, additional Sr-90 will incorporate into the apatite precipitate. Therefore, high pressure water jetting is a viable technology to emplace phosphate or apatite in shallow subsurface sediments difficult to emplace by Ca-citrate-PO4 groundwater injections, but further analysis is needed to quantify the relevant areal extent of phosphate deposition (in the 5- to 15-ft distance from injection points) and cause of the high deposition in finer grained sediments.
Characteristics and Significance of Magma Emplacement Horizons, Black Sturgeon Sill, Nipigon, Ontario

Science.gov (United States)

Zieg, M. J.; Hone, S. V.

2017-12-01

Spatial scales strongly control the timescales of processes in igneous intrusions, particularly through the thermal evolution of the magma, which in turn governs the evolution of crystallinity, viscosity, and other important physical and chemical properties of the system. In this study, we have collected a highly detailed data set comprising geochemical (bulk rock composition), textural (size and alignment of plagioclase crystals), and mineralogical (modal abundance) profiles through the central portion of the 250 m thick Black Sturgeon diabase sill. In this data, we have identified characteristic signals in texture (soft and somewhat diffuse chills), composition (reversals in differentiation trends), and mineralogy (olivine accumulations), all coinciding and recurring at roughly 10 meter intervals. Based on these signatures, we are able to map out multiple zones representing discrete pulses of magma that were emplaced sequentially as the intrusion was inflated. Simple thermal calculations suggest that each 10 meters of new crystallization would require repose times on the order of 10-100 years. To build up 250 meters of magma at this rate would only require approximately 250-2500 years, significantly less than the thermal lifetime of the entire sill. The soft chills we observe in the Black Sturgeon sill are therefore consistent with a system that remained warm throughout the emplacement process. Successive pulses were injected into partially crystalline mush, rather than pure liquid (which would result in hybridization) or solid (which would produce sharp hard chills). Episodic emplacement is by now widely recognized as a fundamental process in the formation of large felsic magma chambers; our results suggest that this also may be an important consideration in understanding the evolution of smaller mafic intrusions.
Spent fuel workshop'2002

Energy Technology Data Exchange (ETDEWEB)

Poinssot, Ch

2002-07-01

This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO{sub 2} fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO{sub 2} dissolution determined from electrochemical experiments with {sup 238}Pu doped UO{sub 2} M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO{sub 2} studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with {alpha} doped UO{sub 2} in Boom clay conditions (K. Lemmens), Studies of the behavior of UO{sub 2} / water interfaces under He{sup 2+} beam (C. Corbel), Alpha and gamma radiolysis effects on UO{sub 2} alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines
The Spent Fuel Management in Finland and Modifications of Spent Fuel Storages

International Nuclear Information System (INIS)

Maaranen, Paeivi

2014-01-01

The objective of this presentation is to share the Finnish regulator's (STUK) experiences on regulatory oversight of the enlargement of a spent fuel interim storage. An overview of the current situation of spent fuel management in Finland will also be given. In addition, the planned modifications and requirements set for spent fuel storages due to the Fukushima accident are discussed. In Finland, there are four operating reactors, one under construction and two reactors that have a Council of State's Decision-in-Principle to proceed with the planning and licensing of a new reactor. In Olkiluoto, the two operating ASEA-Atom BWR units and the Areva EPR under construction have a shared interim storage for the spent fuel. The storage was designed and constructed in 1980's. The option for enlarging the storage was foreseen in the original design. Considering three operating units to produce their spent fuel and the final disposal to begin in 2022, extra space in the spent fuel storage is estimated to be needed in around 2014. The operator decided to double the number of the spent fuel pools of the storage and the construction began in 2010. The capacity of the enlarged spent fuel storage is considered to be sufficient for the three Olkiluoto units. The enlargement of the interim storage was included in Olkiluoto NPP 1 and 2 operating license. The licensing of the enlargement was conducted as a major plant modification. The operator needed the approval from STUK to conduct the enlargement. Prior to the construction of this modification, the operator was required to submit the similar documentation as needed for applying for the construction license of a nuclear facility. When conducting changes in an old nuclear facility, the new safety requirements have to be followed. The major challenge in the designing the enlargement of the spent fuel storage was to modify it to withstand a large airplane crash. The operator chose to cover the pools with protecting slabs and also to
State of the art for fabricating and emplacing concrete containers into large horizontal disposal caverns in the french geological repository - 59267

International Nuclear Information System (INIS)

Bosgiraud, Jean-Michel; Guariso, Maurice; Pineau, Francois

2012-01-01

The research and development work presented in this paper was initialized by Andra in 2007. The work necessary for manufacturing and testing a full scale demonstrator is presently implemented. The case story is twofold. The first part is related to the initial development of a high performance concrete formulation used for fabricating concrete storage containers (containing Intermediate Level and Long Lived Waste primary canisters) to be stacked and emplaced into 400-m long concrete lined horizontal disposal vaults (also called cavern), excavated in the Callovo-Oxfordian clay host formation at a 550 to 600-m depth, with an inside diameter of approximately 8-m. The fabrication of the concrete boxes is illustrated. The second part presents the outcome at the end of the detailed design phase, for a system which is now being manufactured (for further test and assembly), for the emplacement of the concrete containers inside the vault. The application was engineered for remote emplacing a pile of 2 concrete containers (the containers are preliminarily stacked in a pile of 2, inside a hot cell, thanks to a ground travelling gantry crane). The emplacement process is justified and the related emplacement synoptic is illustrated. The test campaign is scheduled in 2011-2012. The successful completion of the technical trials is mandatory to confirm the mechanical feasibility of remotely emplacing concrete containers into large horizontal disposal caverns over long distances. The later display of the machinery at work in Andra's showroom will be instrumental for the confidence building process involving the various stakeholders concerned by the public enquiry period (mid-2013) preceding the deep geological repository license application (2014-2015). (authors)
Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

Energy Technology Data Exchange (ETDEWEB)

Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Banfield, James [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Skutnik, Steven [Univ. of Tennessee, Knoxville, TN (United States)

2014-02-01

This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.
Spent fuel management in Japan

International Nuclear Information System (INIS)

Mineo, H.; Nomura, Y.; Sakamoto, K.

1998-01-01

In Japan 52 commercial nuclear power units are now operated, and the total power generation capacity is about 45 GWe. The cumulative amount of spent fuel arising is about 13,500 tU as of March 1997. Spent fuel is reprocessed, and recovered nuclear materials are to be recycled in LWRs and FBRs. In February 1997 short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, backend measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away from reactor sites, considering the increasing amount of spent fuel arising. Research and development on spent fuel storage has been carried out, particularly on dry storage technology. Fundamental studies are also conducted to implement the burnup credit into the criticality safety design of storage and transportation casks. Rokkasho reprocessing plant is being constructed towards its commencement in 2003, and Pu utilization in LWRs will be started in 1999. Research and development of future recycling technology are also continued for the establishment of nuclear fuel cycle based on FBRs and LWRs. (author)
WP EMPLACEMENT CONTROL AND COMMUNICATION EQUIPMENT DESCRIPTIONS

International Nuclear Information System (INIS)

Raczka, N.T.

1997-01-01

The objective and scope of this document are to list and briefly describe the major control and communication equipment necessary for waste package emplacement at the proposed nuclear waste repository at Yucca Mountain. Primary performance characteristics and some specialized design features of the required equipment are explained and summarized in the individual subsections of this document. This task was evaluated in accordance with QAP-2-0 and found not to be quality affecting. Therefore, this document was prepared in accordance with NAP-MG-012. The following control and communication equipment are addressed in this document: (1) Programmable Logic Controllers (PLC's); (2) Leaky Feeder Radio Frequency Communication Equipment; (3) Slotted Microwave guide Communication Equipment; (4) Vision Systems; (5) Radio Control Equipment; and (6) Enclosure Cooling Systems
Safety aspects of dry spent fuel storage and spent fuel management

International Nuclear Information System (INIS)

Botsch, W.; Smalian, S.; Hinterding, P.; Voelzke, H.; Wolff, D.; Kasparek, E.

2014-01-01

The storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Safety aspects like safe enclosure of radioactive materials, safe removal of decay heat, nuclear criticality safety and avoidance of unnecessary radiation exposure must be achieved throughout the storage period. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. In Germany dual purpose casks for SF or HLW are used for safe transportation and interim storage. TUV and BAM, who work as independent experts for the competent authorities, present the storage licensing process including sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields (authors)
Reference waste form, basalts, and ground water systems for waste interaction studies

Energy Technology Data Exchange (ETDEWEB)

Deju, R.A.; Ledgerwood, R.K.; Long, P.E.

1978-09-01

This report summarizes the type of waste form, basalt, and ground water compositions to be used in theoretical and experimental models of the geochemical environment to be simulated in studying a typical basalt repository. Waste forms to be used in the experiments include, and are limited to, glass, supercalcine, and spent unreprocessed fuel. Reference basalts selected for study include the Pomona member and the Umtanum Unit, Shwana Member, of the Columbia River Basalt Group. In addition, a sample of the Basalt International Geochemical Standard (BCR-1) will be used for cross-comparison purposes. The representative water to be used is of a sodium bicarbonate composition as determined from results of analyses of deep ground waters underlying the Hanford Site. 12 figures, 13 tables.
Reference waste form, basalts, and ground water systems for waste interaction studies

International Nuclear Information System (INIS)

Deju, R.A.; Ledgerwood, R.K.; Long, P.E.

1978-09-01

This report summarizes the type of waste form, basalt, and ground water compositions to be used in theoretical and experimental models of the geochemical environment to be simulated in studying a typical basalt repository. Waste forms to be used in the experiments include, and are limited to, glass, supercalcine, and spent unreprocessed fuel. Reference basalts selected for study include the Pomona member and the Umtanum Unit, Shwana Member, of the Columbia River Basalt Group. In addition, a sample of the Basalt International Geochemical Standard (BCR-1) will be used for cross-comparison purposes. The representative water to be used is of a sodium bicarbonate composition as determined from results of analyses of deep ground waters underlying the Hanford Site. 12 figures, 13 tables
The spent fuel safety experiment

International Nuclear Information System (INIS)

Harmms, G.A.; Davis, F.J.; Ford, J.T.

1995-01-01

The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort
Determination of emplacement mechanism of Zafarghand granitoid Pluton (Southeast of Ardestan by using anisotropy of magnetic susceptibility method (AMS

Directory of Open Access Journals (Sweden)

Mahmoud Sadeghian

2017-03-01

Full Text Available Zafarghand granitoid pluton with compositional range from gabbro to granite and early to middle Miocene age cropped out about 35 km of SE Ardestan. This pluton intrudedthe Eocene volcanic and volcanosedimentary rocks of the Urumieh - Dokhtar structuralzone. In this research, for the first time, the emplacement mechanism of Zafarghandgranitoidic pluton method has been investigated using of anisotropy of magneticsusceptibility (AMS. Based on field observations, as well as petrography andinterpretations of magnetic parameters, Zafarghand pluton divided into 5 domains (1A,1B, 2, 3, 4 and 5. Domain 1, in turn, is divided into 1A and 1B. Domains 2 and 4 arelithologically, gabbro to quartzdiorite and have been emplaced first. They have playedas feeder zones. Domains 1A, 1B, 3, and 5 are dominantly granodioritic to graniticcomposition and have been emplaced as a big and low dip magmatic flow (or possibly as a sill. The occurrence of gabbro to quartzdiorite as well as grandiorite, granite andtonalite in the margin borders of the body, are all indication of magma mixing. It isshould be noted that during emplacement of the pluton studied, fractionalcrystallization, magma mixing and crustal contamination contributed to its generationand the evolution as well.
Evaluation of the post-emplacement environment of high level radioactive waste packages at Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Glassley, W.

1989-01-01

Evaluation of the post-emplacement environment around high-level radioactive waste containers is required by federal regulations. The information derived from this evaluation will be used to determine the service performance of the waste containers, the chemical and hydrological conditions that may influence radionuclide release and transport if containers are breached, and retrievability of the waste containers prior to closure of the repository. Laboratory studies, numerical simulations, and field experiments and tests are used to provide data necessary for this evaluation. Results obtained to date demonstrate that the post-emplacement environment in the welded tuff at Yucca Mountain, Nevada maintains relatively benign chemical features (i.e., near neutral pH, low concentrations of dissolved species) for most scenarios. The hydrological environment appears to be one of low flow volume and rates for the expected condition of an unsaturated medium. Emplacement borehole stability will be a function of fracture density and orientation, which may be influenced by microcrack development. Field studies and numerical simulations are in progress that will extend the results of laboratory studies to long time periods. The extent to which chemical, hydrological and mechanical processes can be adequately coupled through numerical simulations remains a matter of concern
Disposal of spent nuclear fuel

International Nuclear Information System (INIS)

1979-12-01

This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste
Safety case for the disposal of spent nuclear fuel at Olkiluoto. Description of the disposal system 2012

International Nuclear Information System (INIS)

2012-12-01

Description of the Disposal System sits within Posiva Oy's Safety Case 'TURVA-2012' report portfolio and has the objective presenting the initial state of the disposal system for the safety case for the disposal of spent nuclear fuel at Olkiluoto, Finland. Disposal system is an entity composed of a repository system and surface environment. The repository system includes the spent nuclear fuel, canister, buffer, backfill, and closure components as well as the host rock. The repository system components have assigned safety functions (except for the spent nuclear fuel) and are subject to requirements. The initial state is presented for each component, and references to the main supporting reports are given to guide the reader for more details. Conditions for each component vary in time and space, due to the time of emplacement and due to the tolerances set for the compositions, geometries and other properties depending on the component. The disposal operation is foreseen to commence ∼ 2020. At the beginning of the postclosure period, around 2120, all the engineered components have been installed and the operation is finalised. The system evolution during the operational phase is discussed in detail in Performance Assessment. The initial state for the host rock is defined to be essentially equal to the baseline conditions prior to starting the construction of the underground characterisation facility ONKALO. For the surface environment, the initial state is the present conditions prevailing. For any other component of the disposal system, the initial state is defined as the state it has when the direct control over that specific part of the system ceases and only limited information can be made available on the subsequent development of conditions in that part of the system or its near field. (orig.)
Safety case for the disposal of spent nuclear fuel at Olkiluoto. Description of the disposal system 2012

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-12-15

Description of the Disposal System sits within Posiva Oy's Safety Case 'TURVA-2012' report portfolio and has the objective presenting the initial state of the disposal system for the safety case for the disposal of spent nuclear fuel at Olkiluoto, Finland. Disposal system is an entity composed of a repository system and surface environment. The repository system includes the spent nuclear fuel, canister, buffer, backfill, and closure components as well as the host rock. The repository system components have assigned safety functions (except for the spent nuclear fuel) and are subject to requirements. The initial state is presented for each component, and references to the main supporting reports are given to guide the reader for more details. Conditions for each component vary in time and space, due to the time of emplacement and due to the tolerances set for the compositions, geometries and other properties depending on the component. The disposal operation is foreseen to commence {approx} 2020. At the beginning of the postclosure period, around 2120, all the engineered components have been installed and the operation is finalised. The system evolution during the operational phase is discussed in detail in Performance Assessment. The initial state for the host rock is defined to be essentially equal to the baseline conditions prior to starting the construction of the underground characterisation facility ONKALO. For the surface environment, the initial state is the present conditions prevailing. For any other component of the disposal system, the initial state is defined as the state it has when the direct control over that specific part of the system ceases and only limited information can be made available on the subsequent development of conditions in that part of the system or its near field. (orig.)
Collective processing device for spent fuel

International Nuclear Information System (INIS)

Irie, Hiroaki; Taniguchi, Noboru.

1996-01-01

The device of the present invention comprises a sealing vessel, a transporting device for transporting spent fuels to the sealing vessel, a laser beam cutting device for cutting the transported spent fuels, a dissolving device for dissolving the cut spent fuels, and a recovering device for recovering radioactive materials from the spent fuels during processing. Reprocessing treatments comprising each processing of dismantling, shearing and dissolving are conducted in the sealing vessel can ensure a sealing barrier for the radioactive materials (fissionable products and heavy nuclides). Then, since spent fuels can be processed in a state of assemblies, and the spent fuels are easily placed in the sealing vessel, operation efficiency is improved, as well as operation cost is saved. Further, since the spent fuels can be cut by a remote laser beam operation, there can be prevented operator's exposure due to radioactive materials released from the spent fuels during cutting operation. (T.M.)

Phase 1 study of metallic cask systems for spent fuel management from reactor to repository. Volume I. Phase 1 study summary

International Nuclear Information System (INIS)

1986-02-01

It was proposed to perform a systems evaluation of metallic cask systems in order to define and examine the use of various metallic cask concepts or combination of concepts for the overall inventory management of spent fuel starting with its discharge from reactors to its emplacement in geologic repositories. This systems evaluation occurs in three phases. This three phase systems evaluation leads to a definition and recommendation of a sound and practical metallic cask system to accomplish efficient and effective management of spent fuel in the back end of the nuclear fuel cycle. Phase 1 Study objectives: establish system-wide functional criteria and assumptions; perform the systems engineering needed to define the metallic cask concepts and their feasibility; perform a screening evaluation of the technical and economic merits of the concepts; and recommend those to be included for a more detailed systems evaluation in Phase 2. Phase 2 Study objectives: refine the system-wide functional criteria and assumptions; perform the design engineering needed to enhance the validity and workability of those concepts recommended in Phase 1; and perform a more detailed systems evaluation. Phase 3 Study objectives: conclude the systems evaluation and develop an implementation plan. Volume I presents an overview of the detailed systems evaluation presented in Volume II
ASSESSMENT OF MICROBIAL LOAD OF SAUSAGES PREPARED FROM DIFFERENT COMBINATION OF SPENT DUCK AND SPENT HEN MEAT

Directory of Open Access Journals (Sweden)

Rajesh Kumar

2016-12-01

Full Text Available Aim of the present study was to assess the microbial load of sausages prepared from different combination of spent duck and spent hen meat. The combination are 100% spent duck (T1, 75%+ 25% spent duck and spent hen (T2, 50%+50% spent duck and spent hen (T3, 25%+75% spent duck and spent hen (T4 and 100% spent hen (T5. All the samples of different combination were subjected to total plate count (TPC, total psychrophilic count (TPSC and total Coliform count (TCC. Mean of TPC for T1, T2, T3, T4 and T5 were 4.69, 4.62, 4.60, 4.49 and 4.46 log 10 CFU/gm respectively, while mean TPSC were 4.46, 4.46, 4.43, 4.36 and 4.36 log CFU/gm respectively There were no significant (p<0.05 difference between the different group of combination of sausages for TPS as well as TPSC but varies significantly (p<0.05 from 14th day of storage in both cases. The coliform group of bacteria will not be detected in any combination of sausages. It is concluded that microbial load of sausage prepared from spent duck is high and it is decreases as the percentage of duck meat decreases but, the upper limit of bacteria in each group of sausages is within limit and hence it is safe for human consumption.
Far-field thermomechanical response of argillaceous rock to emplacement of a nuclear-waste repository

International Nuclear Information System (INIS)

McVey, D.F.; Thomas, R.K.; Lappin, A.R.

1980-08-01

Before heat-producing wastes can be emplaced safely in any argillaceous rock, it will be necessary to understand the far-field thermal and thermomechanical response of this rock to waste emplacement. This report presents the results of a first series of calculations aimed at estimating the far-field response of argillite to waste emplacement. Because the thermal and mechanical properties of argillite are affected by its content of expandable clay, its behavior is briefly compared and contrasted with that of a shale having the same matrix thermal properties, but containing no expandable clay. Under this assumption, modeled temperatures are the same for the two rock types at equivalent power densities and reflect the large dependence of in-situ temperatures on both initial power density and waste type. Thermomechanical calculations indicate that inclusion of contraction behavior of expandable clays in the assumed argillite thermal expansion behavior results, in some cases, in generation of a large zone in and near the repository that has undergone volumetric contraction but is surrounded by uniformly compressive stresses. Information available to date indicates that this contraction would likely result in locally increased fluid permeability and decreased in-situ thermal conductivity, but might well be advantageous as regards radionuclide retention, because of the increased surface area within the contracted zone. Assumption of continuous and positive expansion behavior for the shale eliminates the near-repository contraction and tensional zones, but results in near-surface tensional zones directly above the repository
Probability of spent fuel transportation accidents

International Nuclear Information System (INIS)

McClure, J.D.

1981-07-01

The transported volume of spent fuel, incident/accident experience and accident environment probabilities were reviewed in order to provide an estimate of spent fuel accident probabilities. In particular, the accident review assessed the accident experience for large casks of the type that could transport spent (irradiated) nuclear fuel. This review determined that since 1971, the beginning of official US Department of Transportation record keeping for accidents/incidents, there has been one spent fuel transportation accident. This information, coupled with estimated annual shipping volumes for spent fuel, indicated an estimated annual probability of a spent fuel transport accident of 5 x 10 -7 spent fuel accidents per mile. This is consistent with ordinary truck accident rates. A comparison of accident environments and regulatory test environments suggests that the probability of truck accidents exceeding regulatory test for impact is approximately 10 -9 /mile
Monitoring during the stepwise implementation of the Swedish deep repository for spent fuel

Energy Technology Data Exchange (ETDEWEB)

Baeckblom, Goeran [Conrox (Sweden); Almen, Karl-Erik [KEA Geokonsult AB (Sweden)

2004-03-01

Monitoring in this report is defined as 'Continuous or repeated observations or measurements of parameters to increase the scientific understanding of the site and the repository, to show compliance with requirements or for adaptation of plans in light of the monitoring results.' The international outlook from IAEA, OECD/NEA, CEC and some country-specific reviews presented in the report forms a necessary background to the Swedish monitoring framework. The implementation of the deep repository in Sweden is executed in phases where monitoring is an inherently integrated activity in the programme. The first phase is the site investigations when Primary Baseline conditions are established. During the following construction phase of the repository, detailed site characterisation continues in conjunction with construction of the access to the deposition area, construction of parts of the deposition area and the central service area. Monitoring is then used to track the changes to the previously established Primary Baseline conditions and distinguishing these imposed changes from natural variations or from other man-made influences. During the initial operation phase, around 200-400 canisters of spent fuel is emplaced and deposition tunnels backfilled. After up-dated evaluations, the phase of regular operation begin, where detailed characterisation, construction of the repository and waste emplacement are concurrent activities. The closure of the repository will take place when all spent fuel has been emplaced, i.e. in the latter part of this century. Monitoring during the stepwise implementation of the repository is executed of several reasons mainly to: describe the Primary Baseline conditions of the repository site, develop and demonstrate understanding of the repository site and the behaviour of engineered barriers, assist in the decision-making process, show compliance with international and national guidelines and regulations. Specific rationales for
Spent fuel management overview: a global perspective

International Nuclear Information System (INIS)

Bonne, A.; Crijns, M.J.; Dyck, P.H.; Fukuda, K.; Mourogov, V.M.

1999-01-01

The paper defines the main spent fuel management strategies and options, highlights the challenges for spent fuel storage and gives an overview of the regional balances of spent fuel storage capacity and spent fuel arising. The relevant IAEA activities in the area of spent fuel management are summarised. (author)
Spent nuclear fuel storage - Basic concept

International Nuclear Information System (INIS)

Krempel, Ascanio; Santos, Cicero D. Pacifici dos; Sato, Heitor Hitoshi; Magalhaes, Leonardo de

2009-01-01

According to the procedures adopted in others countries in the world, the spent nuclear fuel elements burned to produce electrical energy in the Brazilian Nuclear Power Plant of Angra do Reis, Central Nuclear Almirante Alvaro Alberto - CNAAA will be stored for a long time. Such procedure will allow the next generation to decide how they will handle those materials. In the future, the reprocessing of the nuclear fuel assemblies could be a good solution in order to have additional energy resource and also to decrease the volume of discarded materials. This decision will be done in the future according to the new studies and investigations that are being studied around the world. The present proposal to handle the nuclear spent fuel is to storage it for a long period of time, under institutional control. Therefore, the aim of this paper is to introduce a proposal of a basic concept of spent fuel storage, which involves the construction of a new storage building at site, in order to increase the present storage capacity of spent fuel assemblies in CNAAA installation; the concept of the spent fuel transportation casks that will transfer the spent fuel assemblies from the power plants to the Spent Fuel Complementary Storage Building and later on from this building to the Long Term Intermediate Storage of Spent Fuel; the concept of the spent fuel canister and finally the basic concept of the spent fuel long term storage. (author)
HFIR spent fuel management alternatives

International Nuclear Information System (INIS)

Begovich, J.M.; Green, V.M.; Shappert, L.B.; Lotts, A.L.

1992-01-01

The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems' Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere
The role of the geothermal gradient in the emplacement and replenishment of ground ice on Mars

Science.gov (United States)

Clifford, Stephen M.

1993-01-01

Knowledge of the mechanisms by which ground ice is emplaced, removed, and potentially replenished, are critical to understanding the climatic and hydrologic behavior of water on Mars, as well as the morphologic evolution of its surface. Because of the strong temperature dependence of the saturated vapor pressure of H2O, the atmospheric emplacement or replenishment of ground ice is prohibited below the depth at which crustal temperatures begin to monotonically increase due to geothermal heating. In contrast, the emplacement and replenishment of ground ice from reservoirs of H2O residing deep within the crust can occur by at least three different thermally-driven processes, involving all three phases of water. In this regard, Clifford has discussed how the presence of a geothermal gradient as small as 15 K/km can give rise to a corresponding vapor pressure gradient sufficient to drive the vertical transport of 1 km of water from a reservoir of ground water at depth to the base of the cryosphere every 10(exp 6) - 10(exp 7) years. This abstract expands on this earlier treatment by considering the influence of thermal gradients on the transport of H2O at temperatures below the freezing point.
Postclosure performance assessment of the SCP [Site Characterization Plan] conceptual design for horizontal emplacement: Revision 1

International Nuclear Information System (INIS)

1987-08-01

This report is a preliminary postclosure performance assessment of the repository design specified in the Site Characterization Plan Conceptual Design Report (SCP-CDR) for horizontal emplacement of high-level nuclear waste. At the time that these analyses were done, horizontal emplacement was the preferred orientation for the waste packages but vertical emplacement is now the reference design. This assessment consists of (1) a review of the regulatory requirements and strategy to demonstrate compliance with these requirements, (2) an analysis of the performance of the total repository system, (3) an analysis of the thermomechanical behavior of the repository, (4) an analysis of brine mobility in the repository, (5) an analysis of the waste package performance, (6) an analysis of the performance of seals, and (7) comments on the sensitivity of the various performance measures to uncertainties in the data and models. These are preliminary analyses and, in most cases, involve bounding calculations of the repository behavior. They have several purposes including (1) assessing how well this conceptual design ''measures up'' against requirements, (2) gaining experience in implementing the performance assessment strategy and tools and thereby learning where improvements are needed, (3) helping to identify needed data, and (4) helping to indicate required design modifications. 26 refs., 40 figs., 20 tabs
Spent fuel treatment in Japan

International Nuclear Information System (INIS)

Takahashi, K.

1999-01-01

In Japan, 52 nuclear power reactors are operating with a total power generation capacity of 45 GWe. The cumulative amount of spent fuel arising, as of March 1998, is about 14,700 W. Spent fuel is reprocessed and recovered nuclear materials are to be recycled in LWRs and FBRs. Pu utilization in LWRs will commence in 1999. In January 1997, short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of the reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, back-end measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away-from-reactor sites, considering the increasing amount of spent fuel arising. Valuable experience was been accumulated at the Tokai Reprocessing Plant (TRP), from the start of hot operation in 1977 up to now. The role of the TRP will be changed from an operation-oriented to a more R and D oriented facility, when PNC is reorganized into the new organization JNC. The Rokkasho reprocessing plant is under construction and is expected to commence operation in 2003. R and D of future recycling technologies is also continued for the establishment of a nuclear fuel cycle based on FBRs and LWRs. (author)
Centralized disassembly and packaging of spent fuel in the DOE spent fuel management system

International Nuclear Information System (INIS)

Johnson, E.R.

1986-01-01

In October 1984, E.R. Johnson Associates, Inc. (JAI) initiated a study of the prospective use of a centralized facility for the disassembly and packaging of spent fuel to support the various elements of the US Dept. of Energy (DOE) spent fuel management system, including facilities for monitored retrievable storage (MRS) and repositories. It was DOE's original plan to receive spent fuel at each repository where it would be disassembled and packaged (overpacked) for disposal purposes. Subsequently, DOE considered the prospective use of MRS of spent fuel as an option for providing safe and reliable management of spent fuel. This study was designed to consider possible advantages of the use of centralized facilities for disassembly and packaging of spent fuel at whose location storage facilities could be added as required. The study was divided into three principal technical tasks that covered: (a) development of requirements and criteria for the central disassembly and packaging facility and associated systems. (2) Development of conceptual designs for the central disassembly and packaging facility and associated systems. (3) Estimation of capital and operating costs involved for all system facilities and determination of life cycle costs for various scenarios of operation - for comparison with the reference system
Final environmental impact statement: US Spent Fuel Policy. Charge for spent fuel storage

International Nuclear Information System (INIS)

1980-05-01

The United States Government policy relating to nuclear fuel reprocessing, which was announced by President Carter on April 7, 1977, provides for an indefinite deferral of reprocessing, and thus commits light water reactor (LWR) plants to a once-through fuel cycle during that indefinite period. In a subsequent action implementing that policy, the Department of Energy (DOE) on October 18, 1977 announced a spent fuel policy which would enable domestic, and on a selective basis, foreign utilities to deliver spent fuel to the US Government for interim storage and final geologic disposal, and pay the Government a fee for such services. This volume addresses itself to whether the fee charged for these services, by its level or its structure, would have any effect on the environmental impacts of implementing the Spent Fuel Policy itself. This volume thus analyzes the fee and various alternatives to determine the interaction between the fee and the degree of participation by domestic utilities and foreign countries in the proposed spent fuel program for implementing the Spent Fuel Policy. It also analyzes the effect, if any, of the fee on the growth of nuclear power
Using restored cross sections to evaluate magma emplacement, White Horse Mountains, Eastern Nevada, U.S.A.

Science.gov (United States)

Marko, Wayne T.; Yoshinobu, Aaron S.

2011-03-01

New field observations and cross section restoration from the Jurassic White Horse pluton-host rock system, Goshute Range, eastern Nevada, USA, indicate a sequential variation of host rock rheology attending magma emplacement. The pluton intruded weakly to nondeformed Devonian-Mississippian limestone, argillite and quartzite at shallow crustal levels (ca. 7 km). The contact aureole is well exposed along the southern, eastern and northern margin of the intrusive body and is less than 1 km wide. Rocks outside of the aureole are sub-horizontal and do not contain a penetrative fabric or are gently folded (interlimb angles > 120°) about sub-vertical axial planes. Within the contact aureole, continuous and discontinuous spaced, axial planar foliations and harmonic to disharmonic, tight to isoclinal folds wrap around the eastern margin of the pluton. Folds verge toward and away from the pluton and rim anticlines, synclines, and monoclines with wavelength in excess of 250 m are preserved along the pluton margin. The spatial proximity of these ductile structures to the pluton and the apparent increase in intensity of structure development approaching the pluton is compatible with contraction within the aureole attending pluton emplacement. However, all of the above structures are truncated by the intrusive contact at various scales. Granodioritic dikes ranging in thickness from 1 m up to ˜ 10 m emanate from the intrusion and cut host rock structure at high angles and turn to propagate towards one another, parallel to the pluton margin and host rock anisotropy. Such features are interpreted to reflect the last stages of diking and brittle deformation that modified the pluton contact after emplacement-related folding of the carbonate rocks, but before final solidification of the pluton. Eight serial geologic cross sections were constructed and evaluated to place geometric constraints on the shape and growth of the White Horse intrusion. Based on line-length restoration of
Spent fuel reprocessing options

International Nuclear Information System (INIS)

2008-08-01

The objective of this publication is to provide an update on the latest developments in nuclear reprocessing technologies in the light of new developments on the global nuclear scene. The background information on spent fuel reprocessing is provided in Section One. Substantial global growth of nuclear electricity generation is expected to occur during this century, in response to environmental issues and to assure the sustainability of the electrical energy supply in both industrial and less-developed countries. This growth carries with it an increasing responsibility to ensure that nuclear fuel cycle technologies are used only for peaceful purposes. In Section Two, an overview of the options for spent fuel reprocessing and their level of development are provided. A number of options exist for the treatment of spent fuel. Some, including those that avoid separation of a pure plutonium stream, are at an advanced level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. In Section Three, research and development in support of advanced reprocessing options is described. Next-generation spent fuel reprocessing plants are likely to be based on aqueous extraction processes that can be designed to a country specific set of spent fuel partitioning criteria for recycling of fissile materials to advanced light water reactors or fast spectrum reactors. The physical design of these plants must incorporate effective means for materials accountancy, safeguards and physical protection. Section four deals with issues and challenges related to spent fuel reprocessing. The spent fuel reprocessing options assessment of economics, proliferation resistance, and environmental impact are discussed. The importance of public acceptance for a reprocessing strategy is discussed. A review of modelling tools to support the
Spent fuel receipt and lag storage facility for the spent fuel handling and packaging program

International Nuclear Information System (INIS)

Black, J.E.; King, F.D.

1979-01-01

Savannah River Laboratory (SRL) is participating in the Spent Fuel Handling and Packaging Program for retrievable, near-surface storage of spent light water reactor (LWR) fuel. One of SRL's responsibilities is to provide a technical description of the wet fuel receipt and lag storage part of the Spent Fuel Handling and Packaging (SFHP) facility. This document is the required technical description
Costing of spent nuclear fuel storage

International Nuclear Information System (INIS)

2009-01-01

This report deals with economic analysis and cost estimation, based on exploration of relevant issues, including a survey of analytical tools for assessment and updated information on the market and financial issues associated with spent fuel storage. The development of new storage technologies and changes in some of the circumstances affecting the costs of spent fuel storage are also incorporated. This report aims to provide comprehensive information on spent fuel storage costs to engineers and nuclear professionals as well as other stakeholders in the nuclear industry. This report is meant to provide informative guidance on economic aspects involved in selecting a spent fuel storage system, including basic methods of analysis and cost data for project evaluation and comparison of storage options, together with financial and business aspects associated with spent fuel storage. After the review of technical options for spent fuel storage in Section 2, cost categories and components involved in the lifecycle of a storage facility are identified in Section 3 and factors affecting costs of spent fuel storage are then reviewed in the Section 4. Methods for cost estimation and analysis are introduced in Section 5, and other financial and business aspects associated with spent fuel storage are discussed in Section 6.
Disposal of spent fuel

International Nuclear Information System (INIS)

Blomeke, J.O.; Ferguson, D.E.; Croff, A.G.

1978-01-01

Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed between the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom
Surveillance instrumentation for spent-fuel safeguards

International Nuclear Information System (INIS)

McKenzie, J.M.; Holmes, J.P.; Gillman, L.K.; Schmitz, J.A.; McDaniel, P.J.

1978-01-01

The movement, in a facility, of spent reactor fuel may be tracked using simple instrumentation together with a real time unfolding algorithm. Experimental measurements, from multiple radiation monitors and crane weight and position monitors, were obtained during spent fuel movements at the G.E. Morris Spent-Fuel Storage Facility. These data and a preliminary version of an unfolding algorithm were used to estimate the position of the centroid and the magnitude of the spent fuel radiation source. Spatial location was estimated to +-1.5 m and source magnitude to +-10% of their true values. Application of this surveillance instrumentation to spent-fuel safeguards is discussed
Preliminary uncertainty analysis of pre-waste-emplacement groundwater travel times for a proposed repository in basalt

International Nuclear Information System (INIS)

Clifton, P.M.; Arnett, R.C.

1984-01-01

Preliminary uncertainty analyses of pre-waste-emplacement groundwater travel times are presented for a potential high-level nuclear waste repository in the deep basalts beneath the Hanford Site, Washington State. The uncertainty analyses are carried out by means of a Monte Carlo technique, which requires the uncertain inputs to be described as either random variables or spatial stochastic processes. Pre-waste-emplacement groundwater travel times are modeled in a continuous, flat-lying basalt flow top that is assumed to overlie the repository horizon. Two-dimensional, steady state groundwater flow is assumed, and transmissivity, effective thickness, and regional hydraulic gradient are considered as uncertain inputs. Groundwater travel time distributions corresponding to three groundwater models are presented and compared. Limitations of these preliminary simulation results are discussed in detail

Spent fuel: prediction model development

International Nuclear Information System (INIS)

Almassy, M.Y.; Bosi, D.M.; Cantley, D.A.

1979-07-01

The need for spent fuel disposal performance modeling stems from a requirement to assess the risks involved with deep geologic disposal of spent fuel, and to support licensing and public acceptance of spent fuel repositories. Through the balanced program of analysis, diagnostic testing, and disposal demonstration tests, highlighted in this presentation, the goal of defining risks and of quantifying fuel performance during long-term disposal can be attained
Waste package performance analysis

International Nuclear Information System (INIS)

Lester, D.H.; Stula, R.T.; Kirstein, B.E.

1982-01-01

A performance assessment model for multiple barrier packages containing unreprocessed spent fuel has been applied to several package designs. The resulting preliminary assessments were intended for use in making decisions about package development programs. A computer model called BARIER estimates the package life and subsequent rate of release of selected nuclides. The model accounts for temperature, pressure (and resulting stresses), bulk and localized corrosion, and nuclide retardation by the backfill after water intrusion into the waste form. The assessment model assumes a post-closure, flooded, geologic repository. Calculations indicated that, within the bounds of model assumptions, packages could last for several hundred years. Intact backfills of appropriate design may be capable of nuclide release delay times on the order of 10 7 yr for uranium, plutonium, and americium. 8 references, 6 figures, 9 tables
Spent fuel storage for ISER plant

International Nuclear Information System (INIS)

Nakajima, Takasuke; Kimura, Yuzi

1987-01-01

ISER is an intrinsically safe reactor basing its safety only on physical laws, and uses a steel reactor vessel in order to be economical. For such a new type reactor, it is essentially important to be accepted by the society by showing that the reactor is more profitable than conventional reactors to the public in both technical and economic viewpoint. It is also important that the reactor raises no serious problem in the total fuel cycle. Reprocessing seems one of the major worldwide fuel cycle issues. Spent fuel storage is also one of the key technologies for fuel cycle back end. Various systems for ISER spent fuel storages are examined in the present report. Spent fuel specifications of ISER are similar to those of LWR and therefore, most of LWR spent fuel technologies are basically applicable to ISER spent fuel. Design requirements and examples of storage facilities are also discussed. Dry storage seems to be preferable for the relatively long cooling time spent fuel like ISER's one from economical viewpoint. Vault storage will possibly be the most advantageous for large storage capacity. Another point for discussion is the location and international collaboration for spent fuel storages: ISER expected to be a worldwide energy source and therefore, international spent fuel management seems to be fairly attractive way for an energy recipient country. (Nogami, K.)
QUALITY AND SHELF LIFE EVALUATION OF NUGGETS PREPARED FROM SPENT DUCK AND SPENT HEN MEAT

Directory of Open Access Journals (Sweden)

Rajesh Kumar

2015-12-01

Full Text Available A study was conducted to compare the quality of nuggets prepared from spent hen and duck meat. The cooked nuggets were analyzed for pH, thiobarbituric acid (TBA, tyrosine value (TV, moisture, fat, protein, total plate count (TPC and sensory evaluations. Nuggets prepared from spent hen meat showed significantly higher (p<0.05 moisture content however pH, fat and protein content were significantly higher (p<0.05 in duck nuggets. TBA values, TVs and (TPC were highest in duck nuggets but were within the acceptable level up to 7th day of refrigerated storage (4±1°C in both types of nuggets. Both nuggets maintain their sensory quality up to 7th day of refrigeration storage but spent hen nuggets were preferred by consumers compared to nuggets prepared from spent duck meat. Result of the study indicated that, despite the comparative differences among these nuggets, spent duck and hen meat could be used for preparation of nutritionally rich and acceptable nuggets.
Handling of spent nuclear fuel and final storage of nitrified high level reprocessing waste

International Nuclear Information System (INIS)

The following stages of handling and transport of the fuel on its way to final storage are dealt with in the report. 1) The spent nuclear fuel is stored at the power station or in the central fuel storage facility awaiting reprocessing. 2) The fuel is reprocessed, i.e. uranium, plutonium and waste are separated from each other. Reprocessing does not take place in Sweden. The highlevel waste is vitrified and can be sent back to Sweden in the 1990s. 3) Vitrified waste is stored for about 30 years awaiting deposition in the final repository. 4) The waste is encapsulated in highly durable materials to prevent groundwater from coming into contact with the waste glass while the radioactivity of the waste is still high. 5) The canisters are emplaced in a final repository which is built at a depth of 500 m in rock of low permeability. 6) All tunnels and shafts are filled with a mixture of clay and sand of low permeability. A detailed analysis of possible harmful effects resulting from normal acitivties and from conceivable accidents is presented in a special section. (author)
Rock cavern storage of spent fuel

Energy Technology Data Exchange (ETDEWEB)

Cho, Won Jin; Kim, Kyung Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kwon, Sang Ki [Inha University, Incheon (Korea, Republic of)

2015-12-15

The rock cavern storage for spent fuel has been assessed to apply in Korea with reviewing the state of the art of the technologies for surface storage and rock cavern storage of spent fuel. The technical feasibility and economic aspects of the rock cavern storage of spent fuel were also analyzed. A considerable area of flat land isolated from the exterior are needed to meet the requirement for the site of the surface storage facilities. It may, however, not be easy to secure such areas in the mountainous region of Korea. Instead, the spent fuel storage facilities constructed in the rock cavern moderate their demands for the suitable site. As a result, the rock cavern storage is a promising alternative for the storage of spent fuel in the aspect of natural and social environments. The rock cavern storage of spent fuel has several advantages compared with the surface storage, and there is no significant difference on the viewpoint of economy between the two alternatives. In addition, no great technical difficulties are present to apply the rock cavern storage technologies to the storage of domestic spent fuel.
The cost of spent fuel storage

Energy Technology Data Exchange (ETDEWEB)

Ramirez S, J. R.; Palacios H, J. C.; Badillo, V.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2011-11-15

Spent fuel is one of the most important issues in the nuclear industry, currently spent fuel management is been cause of great amount of research, investments, constructing repositories or constructing the necessary facilities to reprocess the fuel, and later to recycle the plutonium recovered in thermal reactors. What is the best solution?, or What is the best technology for an specific solution? Many countries have deferred the decision on selecting an option, while others works actively constructing repositories and others implementing the reprocessing facilities to recycle the plutonium obtained from nuclear spent fuel. In Mexico the nuclear power is limited to two reactors BWR type and medium size. So the nuclear spent fuel discharged has been accommodated at reactor's spent fuel pools. Originally these pools have enough capacity to accommodate spent fuel for the 40 years of designed plant operation. However currently, the plants are under a process for extended power up-rate to 20% of original power and also there are plans to extended operational life for 20 more years. Under these conditions there will not be enough room for spent fuel in the pools. (Author)
Transportation of spent nuclear fuels

International Nuclear Information System (INIS)

Meguro, Toshiichi

1976-01-01

The spent nuclear fuel taken out of reactors is cooled in the cooling pool in each power station for a definite time, then transported to a reprocessing plant. At present, there is no reprocessing plant in Japan, therefore the spent nuclear fuel is shipped abroad. In this paper, the experiences and the present situation in Japan are described on the transport of the spent nuclear fuel from light water reactors, centering around the works in Tsuruga Power Station, Japan Atomic Power Co. The spent nuclear fuel in Tsuruga Power Station was first transported in Apr. 1973, and since then, about 36 tons were shipped to Britain by 5 times of transport. The reprocessing plant in Japan is expected to start operation in Apr. 1977, accordingly the spent nuclear fuel used for the trial will be transported in Japan in the latter half of this year. Among the permission and approval required for the transport of spent nuclear fuel, the acquisition of the certificate for transport casks and the approval of land and sea transports are main tasks. The relevant laws are the law concerning the regulations of nuclear raw material, nuclear fuel and reactors and the law concerning the safety of ships. The casks used in Tsuruga Power Station and EXL III type, and the charging of spent nuclear fuel, the decontamination of the casks, the leak test, land transport with a self-running vehicle, loading on board an exclusive carrier and sea transport are briefly explained. The casks and the ship for domestic transport are being prepared. (Kato, I.)
Dry spent fuel storage licensing

International Nuclear Information System (INIS)

Sturz, F.C.

1995-01-01

In the US, at-reactor-site dry spent fuel storage in independent spent fuel storage installations (ISFSI) has become the principal option for utilities needing storage capacity outside of the reactor spent fuel pools. Delays in the geologic repository operational date at or beyond 2010, and the increasing uncertainty of the US Department of Energy's (DOE) being able to site and license a Monitored Retrievable Storage (MRS) facility by 1998 make at-reactor-site dry storage of spent nuclear fuel increasingly desirable to utilities and DOE to meet the need for additional spent fuel storage capacity until disposal, in a repository, is available. The past year has been another busy year for dry spent fuel storage licensing. The licensing staff has been reviewing 7 applications and 12 amendment requests, as well as participating in inspection-related activities. The authors have licensed, on a site-specific basis, a variety of dry technologies (cask, module, and vault). By using certified designs, site-specific licensing is no longer required. Another new cask has been certified. They have received one new application for cask certification and two amendments to a certified cask design. As they stand on the brink of receiving multiple applications from DOE for the MPC, they are preparing to meet the needs of this national program. With the range of technical and licensing options available to utilities, the authors believe that utilities can meet their need for additional spent fuel storage capacity for essentially all reactor sites through the next decade
Development of spent fuel remote handling technology

International Nuclear Information System (INIS)

Yoon, Ji Sup; Park, B. S.; Park, Y. S.; Oh, S. C.; Kim, S. H.; Cho, M. W.; Hong, D. H.

1997-12-01

Since the nation's policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs
Scoping calculations for canister-tunnel migration of corrodants, oxidants and radionuclides

International Nuclear Information System (INIS)

Shaw, W.; Worth, D.

1992-03-01

This report presents the mathematical models and results obtained for a set of scooping calculations which estimate the possible extent of the vertical migration of canister corrodants, oxidants (forming a redox front) and radionuclides between a copper canister containing spent nuclear fuel, and an overlying emplacement tunnel. The KBS-3 concept for the disposal of spent nuclear fuel is copper canisters, vertically emplaced in deposition holes bored in the floor of a tunnel, situated deep underground. The deposition holes are filled with a buffer of bentonite and the tunnel is backfilled with a mixture of sand and bentonite. (au)
Development of spent fuel remote handling technology

Energy Technology Data Exchange (ETDEWEB)

Yoon, Ji Sup; Park, B S; Park, Y S; Oh, S C; Kim, S H; Cho, M W; Hong, D H

1997-12-01

Since the nation`s policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs
Nuclear criticality safety studies applicable to spent fuel shipping cask designs and spent fuel storage

International Nuclear Information System (INIS)

Tang, J.S.

1980-11-01

Criticality analyses of water-moderated and reflected arrays of LWR fresh and spent fuel assemblies were carried out in this study. The calculated results indicate that using the assumption of fresh fuel loading in spent fuel shipping cask design leads to assembly spacings which are about twice the spacings of spent fuel loadings. Some shipping cask walls of composite lead and water are more effective neutron reflectors than water of 30.48 cm
UCRL--84153

International Nuclear Information System (INIS)

Altenbach, T.J.; Lowry, W.E.

1980-01-01

A three-dimensional thermal analysis has been performed using finite difference techniques to determine the near-field response of a baseline spent fuel repository in a deep geologic salt medium. A baseline design incorporates previous thermal modeling experience and OWI recommendations for areal thermal loading in specifying the waste form properties, package details, and emplacement configuration. The base case in this thermal analysis considers one 10-year old PWR spent fuel assembly emplaced to yield a 36 kw/acre (8.9 w/m 2 ) loading. A unit cell model in an infinite array is used to simplify the problem and provide upper-bound temperatures. Boundary conditions are imposed which allow simulations to 1000 years. Variations studied include a comparison of ventilated and unventilated storage room conditions, emplacement packages with and without air gaps surrounding the canister, and room cool-down scenarios with ventilation following an unventilated state for retrieval purposes. At this low power level ventilating the emplacement room has an immediate cooling influence on the canister and effectively maintains the emplacement room floor near the temperature of the ventilating air. The annular gap separating the canister and sleeve causes the peak temperature of the canister surface to rise by 10 0 F (5.6 0 C) over that from a no gap case assuming perfect thermal contact. It was also shown that the time required for the emplacement room to cool down to 1000 0 C) from an unventilated state ranged from 2 weeks to 6 months; when ventilation initiated after times of 5 years to 50 years, respectively. As the work was performed for the Nuclear Regulatory Commission, these results provide a significant addition to the regulatory data base for spent fuel performance in a geological repository
Spent fuels transportation coming from Australia

International Nuclear Information System (INIS)

2002-01-01

Maritime transportation of spent fuels from Australia to France fits into the contract between COGEMA and ANSTO, signed in 1999. This document proposes nine information cards in this domain: HIFAR a key tool of the nuclear, scientific and technological australian program; a presentation of the ANSTO Australian Nuclear Science and Technology Organization; the HIFAR spent fuel management problem; the COGEMA expertise in favor of the research reactor spent fuel; the spent fuel reprocessing at La Hague; the transports management; the transport safety (2 cards); the regulatory framework of the transports. (A.L.B.)
Safety analysis of spent fuel packaging

International Nuclear Information System (INIS)

Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

1987-01-01

Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)
Spent nuclear fuel disposal liability insurance

International Nuclear Information System (INIS)

Martin, D.W.

1984-01-01

This thesis examines the social efficiency of nuclear power when the risks of accidental releases of spent fuel radionuclides from a spent fuel disposal facility are considered. The analysis consists of two major parts. First, a theoretical economic model of the use of nuclear power including the risks associated with releases of radionuclides from a disposal facility is developed. Second, the costs of nuclear power, including the risks associated with a radionuclide release, are empirically compared to the costs of fossil fuel-fired generation of electricity. Under the provisions of the Nuclear Waste Policy Act of 1982, the federally owned and operated spent nuclear fuel disposal facility is not required to maintain a reserve fund to cover damages from an accidental radionuclide release. Thus, the risks of a harmful radionuclide release are not included in the spent nuclear fuel disposal fee charged to the electric utilities. Since the electric utilities do not pay the full, social costs of spent fuel disposal, they use nuclear fuel in excess of the social optimum. An insurance mechanism is proposed to internalize the risks associated with spent fueled disposal. Under this proposal, the Federal government is required to insure the disposal facility against any liabilities arising from accidental releases of spent fuel radionuclides
Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

International Nuclear Information System (INIS)

Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

1982-01-01

The age of nuclear waste - the length of time between its removal from the reactor cores and its emplacement in a repository - is a significant factor in determining the thermal loading of a repository. The surface cooling period as well as the density and sequence of waste emplacement affects both the near-field repository structure and the far-field geologic environment. To investigate these issues, a comprehensive review was made of the available literature pertaining to thermal effects and thermal properties of mined geologic repositories. This included a careful evaluation of the effects of different surface cooling periods of the wastes, which is important for understanding the optimal thermal loading of a repository. The results led to a clearer understanding of the importance of surface cooling in evaluating the overall thermal effects of a radioactive waste repository. The principal findings from these investigations are summarized in this paper
Intermodal transportation of spent fuel

International Nuclear Information System (INIS)

Elder, H.K.

1983-09-01

Concepts for transportation of spent fuel in rail casks from nuclear power plant sites with no rail service are under consideration by the US Department of Energy in the Commercial Spent Fuel Management program at the Pacific Northwest Laboratory. This report identifies and evaluates three alternative systems for intermodal transfer of spent fuel: heavy-haul truck to rail, barge to rail, and barge to heavy-haul truck. This report concludes that, with some modifications and provisions for new equipment, existing rail and marine systems can provide a transportation base for the intermodal transfer of spent fuel to federal interim storage facilities. Some needed land transportation support and loading and unloading equipment does not currently exist. There are insufficient shipping casks available at this time, but the industrial capability to meet projected needs appears adequate
Near surface spent fuel storage: environmental issues

International Nuclear Information System (INIS)

Nelson, I.C.; Shipler, D.B.; McKee, R.W.; Glenn, R.D.

1979-01-01

Interim storage of spent fuel appears inevitable because of the lack of reprocessing plants and spent fuel repositories. This paper examines the environmental issues potentially associated with management of spent fuel before disposal or reprocessing in a reference scenario. The radiological impacts of spent fuel storage are limited to low-level releases of noble gases and iodine. Water needed for water basin storage of spent fuel and transportation accidents are considered; the need to minimize the distance travelled is pointed out. Resource commitments for construction of the storage facilities are analyzed

Spent Fuel Management Newsletter. No. 1

International Nuclear Information System (INIS)

1990-03-01

This Newsletter has been prepared in accordance with the recommendations of the International Regular Advisory Group on Spent Fuel Management and the Agency's programme (GC XXXII/837, Table 76, item 14). The main purpose of the Newsletter is to provide Member States with new information about the state-of-the-art in one of the most important parts of the nuclear fuel cycle - Spent Fuel Management. The contents of this publication consists of two parts: (1) IAEA Secretariat contribution -work and programme of the Nuclear Materials and Fuel Cycle Technology Section of the Division of Nuclear Fuel Cycle and Waste Management, recent and planned meetings and publications, Technical Co-operation projects, Co-ordinated Research programmes, etc. (2) Country reports - national programmes on spent fuel management: current and planned storage and reprocessing capacities, spent fuel arisings, safety, transportation, storage, treatment of spent fuel, some aspects of uranium and plutonium recycling, etc. The IAEA expects to publish the Newsletter once every two years between the publications of the Regular Advisory Group on Spent Fuel Management. Figs and tabs
Understanding the mechanical coupling between magma emplacement and the resulting deformation: the example of saucer-shaped sills

Science.gov (United States)

Galland, O.; Neumann, E. R.; Planke, S.

2009-12-01

The mechanical coupling between magma intrusions and the surrounding rocks plays a major role in the emplacement of volcanic plumbing systems. The deformation associated with magma emplacement has been widely studied, such as caldera inflation/deflation, volcano deformation during dike intrusion, and doming above laccoliths. However, the feedback processes, i.e. the effect of deformation resulting from intruding magma on the propagation of the intrusion itself, have rarely been studied. Saucer-shaped sills are adequate geological objects to understand such processes. Indeed, observation show that saucer-shaped sills are often associated with dome-like structures affecting the overlying sediments. In addition, there is a clear geometrical relation between the sills and the domes: the dome diameters are almost identical to those of saucers, and the tips of the inclined sheets of saucers are superimposed on the edges of the domes. In this presentation, we report on experimental investigations of the emplacement mechanisms of saucer-shaped sills and associated deformation. The model materials were (1) cohesive fine-grained silica flour, representing brittle crust, and (2) molten low-viscosity oil, representing magma. A weak layer located at the top of the injection inlet simulates strata. The main variable parameter is injection depth. During experiments, the surface of the model is digitalized through a structured light technique based on the moiré projection principle. Such a tool provides topographic maps of the surface of the model and allows a periodic (every 1.5 s) monitoring of the model topography. When the model magma starts intruding, a symmetrical dome rises above the inlet. Subsequently, the dome inflates and widens, and then evolves to a plateau-like feature, with nearly flat upper surface and steep sides. At the end of the experiments, the intruding liquid erupts at the edge of the plateau. The intrusions formed in the experiments are saucer-shaped sills
Spent Nuclear Fuel project, project management plan

International Nuclear Information System (INIS)

Fuquay, B.J.

1995-01-01

The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project
Fact sheet on spent fuel management

International Nuclear Information System (INIS)

2006-01-01

The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs. The proceedings of the 2003 IAEA conference on storage of spent fuel from power reactors has been ranked in the top twenty most accessed IAEA publications. These proceedings are available for free downloads at http://www-pub.iaea.org/MTCD/publications/PubDetails.asp?pubId=6924]. The IAEA organized and held a 2004 meeting focused on long term spent fuel storage provisions in Central and Eastern Europe, using technical cooperation funds to support participation by these Member States. Over ninety percent of the participants in this meeting rated its value as good or excellent, with participants noting that the IAEA is having a positive effect in stimulating communication, cooperation, and information dissemination on this important topic. The IAEA was advised in 2004 that results from a recent coordinated research project (IAEA-TECDOC-1343) were used by one Member State to justify higher clad temperatures for spent fuel in dry storage, leading to more efficient storage and reduced costs. Long term
High-precision 40Ar/39AR age of the gas emplacement into the Songliao Basin

NARCIS (Netherlands)

Qiu, H.N.; Wu, H.Y.; Yun, Y.B.; Feng, Z.H.; Xu, Y.G.; Mei, L.F.; Wijbrans, J.R.

2011-01-01

The problem of determining an exact isotopic age of hydrocarbon emplacement is complex because minerals suitable for dating with common isotopic methods are often lacking in the sedimentary domain. However, the igneous quartz from the Cretaceous volcanic rocks that host the gas reservoir in the
Spent Fuel in Chile

International Nuclear Information System (INIS)

López Lizana, F.

2015-01-01

The government has made a complete and serious study of many different aspects and possible road maps for nuclear electric power with strong emphasis on safety and energy independence. In the study, the chapter of SFM has not been a relevant issue at this early stage due to the fact that it has been left for later implementation stage. This paper deals with the options Chile might consider in managing its Spent Fuel taking into account foreign experience and factors related to safety, economics, public acceptance and possible novel approaches in spent fuel treatment. The country’s distinctiveness and past experience in this area taking into account that Chile has two research reactors which will have an influence in the design of the Spent Fuel option. (author)
Assessment of spent fuel cooling

International Nuclear Information System (INIS)

Ibarra, J.G.; Jones, W.R.; Lanik, G.F.

1997-01-01

The paper presents the methodology, the findings, and the conclusions of a study that was done by the Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) on loss of spent fuel pool cooling. The study involved an examination of spent fuel pool designs, operating experience, operating practices, and procedures. AEOD's work was augmented in the area of statistics and probabilistic risk assessment by experts from the Idaho Nuclear Engineering Laboratory. Operating experience was integrated into a probabilistic risk assessment to gain insight on the risks from spent fuel pools
Mechanical degradation of Emplacement Drifts at Yucca Mountain - A Modeling Case Study. Part I: Nonlithophysal Rock

International Nuclear Information System (INIS)

M. Lin; D. Kicker; B. Damjanac; M. Board; M. Karakouzian

2006-01-01

This paper outlines rock mechanics investigations associated with mechanical degradation of planned emplacement drifts at Yucca Mountain, which is the designated site for the proposed U.S. high-level nuclear waste repository. The factors leading to drift degradation include stresses from the overburden, stresses induced by the heat released from the emplaced waste, stresses due to seismically related ground motions, and time-dependent strength degradation. The welded tuff emplacement horizon consists of two groups of rock with distinct engineering properties: nonlithophysal units and lithophysal units, based on the relative proportion of lithophysal cavities. The term 'lithophysal' refers to hollow, bubble like cavities in volcanic rock that are surrounded by a porous rim formed by fine-grained alkali feldspar, quartz, and other minerals. Lithophysae are typically a few centimeters to a few decimeters in diameter. Part I of the paper concentrates on the generally hard, strong, and fractured nonlithophysal rock. The degradation behavior of the tunnels in the nonlithophysal rock is controlled by the occurrence of keyblocks. A statistically equivalent fracture model was generated based on extensive underground fracture mapping data from the Exploratory Studies Facility at Yucca Mountain. Three-dimensional distinct block analyses, generated with the fracture patterns randomly selected from the fracture model, were developed with the consideration of in situ, thermal, and seismic loads. In this study, field data, laboratory data, and numerical analyses are well integrated to provide a solution for the unique problem of modeling drift degradation
Spent fuel storage process equipment development

International Nuclear Information System (INIS)

Park, Hyun Soo; Lee, Jae Sol; Yoo, Jae Hyung

1990-02-01

Nuclear energy which is a major energy source of national energy supply entails spent fuels. Spent fuels which are high level radioactive meterials, are tricky to manage and need high technology. The objectives of this study are to establish and develop key elements of spent fuel management technologies: handling equipment and maintenance, process automation technology, colling system, and cleanup system. (author)
Effects of the deviation characteristics of nuclear waste emplacement boreholes on borehole liner stresses

International Nuclear Information System (INIS)

Glowka, D.A.

1990-09-01

This report investigates the effects of borehole deviation on the useability of lined boreholes for the disposal of high-level nuclear waste at the proposed Yucca Mountain Repository in Nevada. Items that lead to constraints on borehole deviation include excessive stresses that could cause liner failure and possible binding of a waste container inside the liner during waste emplacement and retrieval operations. Liner stress models are developed for two general borehole configurations, one for boreholes drilled with a steerable bit and one for boreholes drilled with a non-steerable bit. Procedures are developed for calculating liner stresses that arise both during insertion of the liner into a borehole and during the thermal expansion process that follows waste emplacement. The effects of borehole curvature on the ability of the waste container to pass freely inside the liner without binding are also examined. Based on the results, specifications on borehole deviation allowances are developed for specific vertical and horizontal borehole configurations of current interest. 11 refs., 22 figs., 4 tabs
Spent fuel dry storage in Hungary

International Nuclear Information System (INIS)

Buday, G.; Szabo, B.; Oerdoegh, M.; Takats, F.

1999-01-01

Paks Nuclear Power Plant is the only NPP in Hungary. It has four WWER-440 type reactor units. Since 1989, approximately 40-50% of the total annual electricity generation of the country has been supplied by this plant. The fresh fuel is imported from Russia. Most of the spent fuel assemblies have been shipped back to Russia. Difficulties with spent fuel transportation to Russia have begun in 1992. Since that time, some of the shipments were delayed, some of them were completely cancelled, thus creating a backlog of spent fuel filling all storage positions of the plant. To provide assurance of the continued operation, Paks NPPs management decided to implement an independent spent fuel storage facility and chose GEC-Althom's MVDS design. The construction of the facility started in February 1995 and the first spent fuel assembly was placed in the store in September 1997. The paper gives an overview of the situation, describing the conditions leading to the construction of the dry storage facility at Paks and its implementation. Finally, some information is given about the new Public Agency for Radioactive Waste Management established this year and responsible for managing the issues related to spent fuel management. (author)
Transpressional granite-emplacement model: Structural and magnetic study of the Pan-African Bandja granitic pluton (West Cameroon)

Science.gov (United States)

Sandjo, A. F. Yakeu; Njanko, T.; Njonfang, E.; Errami, E.; Rochette, P.; Fozing, E.

2016-02-01

The Pan-African NE-SW elongated Bandja granitic pluton, located at the western part of the Pan-African belt in Cameroon, is a K-feldspar megacryst granite. It is emplaced in banded gneiss and its NW border underwent mylonitization. The magmatic foliation shows NE-SW and NNE-SSW strike directions with moderate to strong dip respectively in its northern and central parts. This mostly, ferromagnetic granite displays magnetic fabrics carried by magnetite and characterized by (i) magnetic foliation with best poles at 295/34, 283/33 and 35/59 respectively in its northern, central and southern parts and (ii) a subhorizontal magnetic lineation with best line at 37/8, 191/9 and 267/22 respectively in the northern, central and southern parts. Magnetic lineation shows an `S' shape trend that allows to (1) consider the complete emplacement and deformation of the pluton during the Pan-African D 2 and D 3 events which occurred in the Pan-African belt in Cameroon and (2) reorganize Pan-African ages from Nguiessi Tchakam et al. (1997) compared with those of the other granitic plutons in the belt as: 686 ±17 Ma (Rb/Sr) for D 1 age of metamorphism recorded in gneiss; and the period between 604-557 Ma for D 2-D 3 emplacement and deformation age of the granitic pluton in a dextral ENE-WSW shear movement.
Overview of spent fuel management and problems

International Nuclear Information System (INIS)

Ritchie, I.G.; Ernst, P.C.

1998-01-01

Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel worldwide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialized and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. Some projections of spent fuel inventories to the year 2006 are presented and discussed. (author)
Modular dry storage of spent fuel

International Nuclear Information System (INIS)

Baxter, J.W.

1982-01-01

Long term uncertainties in US spent fuel reprocessing and storage policies and programs are forcing the electric utilities to consider means of storing spent fuel at the reactor site in increasing quantitities and for protracted periods. Utilities have taken initial steps in increasing storage capacity. Existing wet storage pools have in many cases been reracked to optimize their capacity for storing spent fuel assemblies
Spent fuel element storage facility

International Nuclear Information System (INIS)

Ukaji, Hideo; Yamashita, Rikuo.

1981-01-01

Purpose: To always keep water level of a spent fuel cask pit equal with water level of spent fuel storage pool by means of syphon principle. Constitution: The pool water of a spent fuel storage pool is airtightly communicated through a pipe with the pool water of a spent fuel cask, and a gate is provided between the pool and the cask. Since cask is conveyed into the cask pit as the gate close while conveying, the pool water level is raised an amount corresponding to the volume of the cask, and water flow through scattering pipe and the communication pipe to the storage pool. When the fuel is conveyed out of the cask, the water level is lowered in the amount corresponding to the volume in the cask pit, and the water in the pool flow through the communication pipe to the cask pit. (Sekiya, K.)
The Influence of Topography on the Emplacement Dynamics of Martian Lava flows

Science.gov (United States)

Tremblay, J.; Fitch, E. P.; Fagents, S. A.

2017-12-01

Lava flows on the Martian surface exhibit a diverse array of complex morphologies. Previous emplacement models, based on terrestrial flows, do not fully account for these observed complex morphologies. We assert that the topography encountered by the flow can exert substantial control over the thermal, rheological, and morphological evolution of the flow, and that these effects can be better incorporated into flow models to predict Martian flow morphologies. Our development of an updated model can be used to account for these topographical effects and better constrain flow parameters. The model predicts that a slope break or flow meander induces eddy currents within the flow, resulting in the disruption of the flow surface crust. The exposure of the flow core results in accelerated cooling of the flow and a resultant increase in viscosity, leading to slowing of the flow. A constant source lava flux and a stagnated flow channel would then result in observable morphological changes, such as overflowing of channel levees. We have identified five morphological types of Martian flows, representing a range of effusion rates, eruption durations and topographic settings, which are suitable for application of our model. To characterize flow morphology, we used imaging and topographic data sets to collect data on flow dimensions. For eight large (50 to hundreds of km long) channelized flows in the Tharsis region, we used the MOLA 128 ppd DEM and/or individual MOLA shot points to derive flow cross-sectional thickness profiles, from which we calculated the cross-sectional area of the flow margins adjacent to the main channel. We found that the largest flow margin cross sectional areas (excluding the channel) occur in association with a channel bend, typically near the bend apex. Analysis of high-resolution images indicates that these widened flow margins are the result of repeated overflows of the channel levees and emplacement of short flow lobes adjacent to the main flow. In
CLASSIFICATION OF THE MGR WASTE EMPLACEMENT/RETRIEVAL SYSTEM

International Nuclear Information System (INIS)

J.A. Ziegler

2000-01-01

The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) waste emplacement/retrieved system structures, systems and components (SSCs) performed by the MGR Preclosure Safety and Systems Engineering Section. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 2000). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, Quality Assurance Requirements and Description (QARD) (DOE 2000). This QA classification incorporates the current MGR design and the results of the ''Design Basis Event Frequency and Dose Calculation for Site Recommendation'' (CRWMS M andO 2000a). The content and technical approach of this analysis is in accordance with the development plan ''QA Classification of MGR Structures, Systems, and Components'' (CRWMS M andO 1999b)
Remote excavation using the telerobotic small emplacement excavator

International Nuclear Information System (INIS)

Thompson, D.H.; Burks, B.L.; Killough, S.M.

1993-01-01

Oak Ridge National Laboratory is developing remote excavation technologies for the Office of Technology Development, Robotics Technology Development Program. This work is being done to meet the need for remote excavation and removal of radioactive and contaminated buried waste at several DOE sites. System requirements are based on the need to uncover and remove waste from burial sites in a way that does not cause unnecessary personnel exposure or additional environmental contamination. Goals for the current project are to demonstrate dexterous control of a backhoe with force feedback and to implement robotic operations that will improve productivity. The Telerobotic Small Emplacement Excavator is a prototype system that incorporates the needed robotic and telerobotic capabilities on a commercially available platform. The ability to add remote dexterous teleoperation and robotic operating modes is intended to be adaptable to other commercially available excavator systems
Relating rock avalanche morphology to emplacement processes

Science.gov (United States)

Dufresne, Anja; Prager, Christoph; Bösmeier, Annette

2015-04-01

The morphology, structure and sedimentological characteristics of rock avalanche deposits reflect both internal emplacement processes and external influences, such as runout path characteristics. The latter is mainly predisposed by topography, substrate types, and hydrogeological conditions. Additionally, the geological setting at the source slope controls, e.g. the spatial distribution of accumulated lithologies and hence material property-related changes in morphology, or the maximum clast size and amount of fines of different lithological units. The Holocene Tschirgant rock avalanche (Tyrol, Austria) resulted from failure of an intensely deformed carbonate rock mass on the southeast face of a 2,370-m-high mountain ridge. The initially sliding rock mass rapidly fragmented as it moved towards the floor of the Inn River valley. Part of the 200-250 x 106 m3 (Patzelt 2012) rock avalanche debris collided with and moved around an opposing bedrock ridge and flowed into the Ötz valley, reaching up to 6.3 km from source. Where the Tschirgant rock avalanche spread freely it formed longitudinal ridges aligned along motion direction as well as smaller hummocks. Encountering high topography, it left runup ridges, fallback patterns (i.e. secondary collapse), and compressional morphology (successively elevated, transverse ridges). Further evidence for the mechanical landslide behaviour is given by large volumes of mobilized valley-fill sediments (polymict gravels and sands). These sediments indicate both shearing and compressional faulting within the rock avalanche mass (forming their own morphological units through, e.g. in situ bulldozing or as distinctly different hummocky terrain), but also indicate extension of the spreading landslide mass (i.e. intercalated/injected gravels encountered mainly in morphological depressions between hummocks). Further influences on its morphology are given by the different lithological units. E.g. the transition from massive dolomite
Geochronological Constraints on the Exhumation and Emplacement of Subcontinental Lithospheric Mantle Peridotites in the Westernmost Mediterranean

Science.gov (United States)

Garrido, Carlos J.; Hidas, Károly; Marchesi, Claudio; Varas-Reus, María Isabel; Booth-Rea, Guillermo

2017-04-01

Exhumation of subcontinental mantle peridotite in the Western Mediterranean has been attributed to different tectonic processes including pure extension, transpression, or alternating contractive and extensional processes related with continental subduction followed by extension, before final their contractive intracrustal emplacement. Any model trying to explain the exhumation and emplacement of subcontinental lithospheric mantle peridotites in the westernmost Mediterranean should take into account the available geochronological constraints, as well as the petrological and geochemical processes that lead to internal tectono-magmatic zoning so characteristic of the Betic and Rif orogenic peridotites. Different studies have suggested a Hercynian, Cenozoic-Mesozoic or an Alpine age for the late tectono-magmatic evolution and intra-crustal emplacement of Betic-Rif peridotites. The pervasive presence of Mesozoic U-Pb zircon ages in Ronda UHP and HP garnet pyroxenites does not support a Hercynian age for the intracrustal emplacement of the peridotite. A hyper-extended margin setting for is in good agreement with the Jurassic extensional event that pervasively affected ALKAPECA terrains (i.e. the Alboran, Kabylides, Peloritani, and Calabria domains) in the western Mediterranean due to the opening of the Piemonte-Ligurian Ocean. However, a Jurassic age and a passive margin tectonic setting do not account, among other observations, for the late Miocene thermochronological ages recorded in zircons rims (U-Pb) and garnets (Lu-Hf) in garnet pyroxenites from the Betic-Rif peridotites, the pervasive Miocene resetting of U-Pb zircon and monazite ages in the overlying Jubrique crustal section, the supra-subduction radiogenic signature of late pyroxenite intrusive dikes in the Ronda peridotite, and the arc tholeiitic affinity of late mantle-derived, gabbroic dykes intruding in the Ronda and Ojen plagioclase lherzolites. These data are more consistent with a supra

Overview on spent fuel management strategies

International Nuclear Information System (INIS)

Dyck, P.

2002-01-01

This paper presents an overview on spent fuel management strategies which range from reprocessing to interim storage in a centralised facility followed by final disposal in a repository. In either case, more spent fuel storage capacity (wet or dry, at-reactor or away-from-reactor, national or regional) is required as spent fuel is continuously accumulated while most countries prefer to defer their decision to choose between these two strategies. (author)
Probable leaching mechanisms for spent fuel

International Nuclear Information System (INIS)

Wang, R.; Katayama, Y.B.

1981-01-01

At the Pacific Northwest Laboratory, researchers in the Waste/Rock Interaction Technology Program are studying spent fuel as a possible waste form for the Office of Nuclear Waste Isolation. This paper presents probable leaching mechanisms for spent fuel and discusses current progress in identifying and understanding the leaching process. During the past year, experiments were begun to study the complex leaching mechanism of spent fuel. The initial work in this investigation was done with UO 2 , which provided the most information possible on the behavior of the spent-fuel matrix without encountering the very high radiation levels associated with spent fuel. Both single-crystal and polycrystalline UO 2 samples were used for this study, and techniques applicable to remote experimentation in a hot cell are being developed. The effects of radiation are being studied in terms of radiolysis of water and surface activation of the UO 2 . Dissolution behavior and kinetics of UO 2 were also investigated by electrochemical measurement techniques. These data will be correlated with those acquired when spent fuel is tested in a hot cell. Oxidation effects represent a major area of concern in evaluating the stability of spent fuel. Dissolution of UO 2 is greatly increased in an oxidizing solution because the dissolution is then controlled by the formation of hexavalent uranium. In solutions containing very low oxygen levels (i.e., reducing solutions), oxidation-induced dissolution may be possible via a previously oxidized surface, through exposure to air during storage, or by local oxidants such as O 2 and H 2 O 2 produced from radiolysis of water and radiation-activated UO 2 surfaces. The effects of oxidation not only increase the dissolution rate, but could lead to the disintegration of spent fuel into fine fragments
Survey of wet and dry spent fuel storage

International Nuclear Information System (INIS)

1999-07-01

Spent fuel storage is one of the important stages in the nuclear fuel cycle and stands among the most vital challenges for countries operating nuclear power plants. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for exchanging information and for coordinating and encouraging closer co-operation among Member States. Spent fuel management is recognized as a high priority IAEA activity. In 1997, the annual spent fuel arising from all types of power reactors worldwide amounted to about 10,500 tonnes heavy metal (t HM). The total amount of spent fuel accumulated worldwide at the end of 1997 was about 200,000 t HM of which about 130,000 t HM of spent fuel is presently being stored in at-reactor (AR) or away-from-reactor (AFR) storage facilities awaiting either reprocessing or final disposal and 70,000 t HM has been reprocessed. Projections indicate that the cumulative amount generated by 2010 may surpass 340,000 t HM and by the year 2015 395,000 t HM. Part of the spent fuel will be reprocessed and some countries took the option to dispose their spent fuel in a repository. Most countries with nuclear programmes are using the deferral of a decision approach, a 'wait and see' strategy with interim storage, which provides the ability to monitor the storage continuously and to retrieve the spent fuel later for either direct disposal or reprocessing. Some countries use different approaches for different types of fuel. Today the worldwide reprocessing capacity is only a fraction of the total spent fuel arising and since no final repository has yet been constructed, there will be an increasing demand for interim storage. The present survey contains information on the basic storage technologies and facility types, experience with wet and dry storage of spent fuel and international experience in spent fuel transport. The main aim is to provide spent fuel
Spent fuel interim management: 1995 update

International Nuclear Information System (INIS)

Anderson, C.K.

1995-01-01

The problems of interim away-from-reactor spent fuel storage and storage in spent fuel pools at the reactor site are discussed. An overview of the state-of-the-art in the USA, Europe, and Japan is presented. The technical facilities for away-from-reactor storage are briefly described, including wet storage pools, interactive concrete systems, metallic containers, and passive concrete systems. Reprocessing technologies are mostly at the design stage only. It is predicted that during the 20 years to come, about 50 000 tonnes of spent fuel will be stored at reactor sites regardless of the advance of spent fuel reprocessing or interim storage projects. (J.B.). 4 tabs., 2 figs
Deformation mechanisms accommodating the emplacement of an igneous sill-complex in the Irish sector of the Rockall Basin, offshore NW Ireland

Science.gov (United States)

Magee, Craig; Jackson, Christopher; Schofield, Nick

2013-04-01

Magma intrusion within the subsurface is heavily influenced by the pre-existing architecture of the upper crust and, depending on the emplacement mechanisms, may modify basin structure and fluid flow pathways. Seismic reflection data reveal that intrusive networks predominantly consist of interconnected, saucer-shaped sills that are often associated with dome-shaped 'forced' folds generated by intrusion-induced uplift. Previous studies of intrusion-related forced folds have primarily focused on isolated sills or laccoliths and have shown that the fold amplitude is less than the intrusion thickness, suggesting that additional space-making mechanisms (e.g. grain comminution, fluidization) accompanied emplacement. Furthermore, forced folding is often considered as an instantaneous process over geological time. However, fold growth and the interplay between accompanying ductile and brittle deformation styles remains poorly understood. Here, we use 3D seismic reflection data from the eastern margin of the Irish Rockall Basin, NE Atlantic, to quantitatively study eighty-two igneous intrusions (i.e. saucer-shaped sills and inclined sheets) in order to constrain the emplacement history of a Palaeocene-to-Middle Eocene sill-complex. Emplacement occurred across a Cretaceous clastic-to-marl dominated succession at palaeodepths of limbs. The saucer-shaped sills are characterized by radial magma flow patterns, emanating from the inner sill, distinguished by mapping the long axes of magma lobes and fingers. These magma flow indictors also provide a proxy for intrusion style; i.e. where sills intrude the Lower Cretaceous sandstones, magma propagation was facilitated by brittle fracturing while non-brittle processes (e.g. fluidization) accompanied shallow-level intrusions into the Upper Cretaceous marls. Directly overlying the sill-complex, within the Palaeocene-Mid-Eocene succession, are a series of 22 forced folds that are interpreted to have formed via intrusion-induced uplift
Spent fuel management in Canada

International Nuclear Information System (INIS)

Khan, A.; Pattantyus, P.

1999-01-01

The current status of the Canadian spent fuel storage is presented. This includes wet and dry interim storage. Extension of wet interim storage facilities is nor planned, as dry technologies have found wide acceptance. The Canadian nuclear program is sustained by commercial Ontario Hydro CANDU type reactors, since 1971, representing 13600 MW(e) of installed capacity, able to produce 9200 spent fuel bundles (1800 tU) every year, and Hydro Quebec and New Brunswick CANDU reactors each producing 685 MW(e) and about 100 tU of spent fuel annually. The implementation of various interim (wt and dry) storage technologies resulted in simple, dense and low cost systems. Economical factors determined that the open cycle option be adopted for the CANDU type reactors rather that recycling the spent fuel. Research and development activities for immobilization and final disposal of nuclear waste are being undertaken in the Canadian Nuclear Fuel Waste Management Program
Development of advanced spent fuel management process. System analysis of advanced spent fuel management process

International Nuclear Information System (INIS)

Ro, S.G.; Kang, D.S.; Seo, C.S.; Lee, H.H.; Shin, Y.J.; Park, S.W.

1999-03-01

The system analysis of an advanced spent fuel management process to establish a non-proliferation model for the long-term spent fuel management is performed by comparing the several dry processes, such as a salt transport process, a lithium process, the IFR process developed in America, and DDP developed in Russia. In our system analysis, the non-proliferation concept is focused on the separation factor between uranium and plutonium and decontamination factors of products in each process, and the non-proliferation model for the long-term spent fuel management has finally been introduced. (Author). 29 refs., 17 tabs., 12 figs
Geology and tectonic magmatic of emplacement of a longitudinal dyke swarm of Nico Perez(Minas) URUGUAY

International Nuclear Information System (INIS)

Gonzalez, P.; Poire, D.; Canalicchio, J.; Garcia Repetto, F.

2004-01-01

The Mina Verdun Group (Precambrian) was deposited prior to the subvolcanic emplacement of a longitudinal dyke swarm of basaltic to andesitic composition (Minas Subvolcanic Swarm of the Mina Verdun quarry - Nico Perez Terrane, Minas, Uruguay). The swarm and its country rocks predated a tectono-metamorphic event that produced fragileductile shear zones associated with very low- to low-grade dislocation metamorphism. We interpreted a K-Ar whole rock datum of 485,2 ± 12,5 Ma (andesitic dyke) as a minimum cooling age in relation with late- to post-swarm emplacement deuteric alteration stage. Another K-Ar whole rock datum of 108,5 ± 2,9 Ma on a basaltic dyke was assumed here as a degasification stage, while its geological meaning is still matter of debate. The Minas Subvolcanic Dyke Swarm was intruded at high crustal levels, suggesting that the Minas region was affected by a period of extensional tectonics [es
Automatic spent fuel ID number reader (I)

International Nuclear Information System (INIS)

Tanabe, S.; Kawamoto, H.; Fujimaki, K.; Kobe, A.

1991-01-01

An effective and efficient technique has been developed for facilitating identification works of LWR spent fuel stored in large scale spent fuel storage pools of such as processing plants. Experience shows that there are often difficulties in the implementation of operator's nuclear material accountancy and control works as well as safeguards inspections conducted on spent fuel assemblies stored in deep water pool. This paper reports that the technique is realized as an automatic spent fuel ID number reader system installed on fuel handling machine. The ID number reader system consists of an optical sub-system and an image processing sub-system. Thousands of spent fuel assemblies stored in under water open racks in each storage pool could be identified within relatively short time (e.g. within several hours) by using this combination. Various performance tests were carried out on image processing sub-system in 1990 using TV images obtained from different types of spent fuel assemblies stored in various storage pools of PWR and BWR power stations
Management plan for the procurement of shipping casks required to service proposed federal waste repositories

International Nuclear Information System (INIS)

Renken, J.H.; Dupree, S.A.; Allen, G.C. Jr.; Freedman, J.M.

1978-08-01

Development of transportation systems to move radioactive waste and unreprocessed spent fuel to proposed federal waste repositories is an integral part of the National Waste Terminal Storage Program. To meet this requirement, shipping casks must be designed, licensed, and fabricated. To assist the manager charged with this responsibility, a Cask Procurement Plan has been formulated. This plan is presented as a logic diagram that is suitable for computer analysis. In addition to the diagram, narrative material that describes various activities in the plan is also included. A preliminary computer analysis of the logic diagram indicates that, depending on the result of several decisions which must be made during the course of the work, the latest start dates which will allow prototype delivery of all types of casks by December 1985, range from November 1977 to March 1982
Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

International Nuclear Information System (INIS)

Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

1988-12-01

This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO 2 oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO 2 pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs
Remote technology applications in spent fuel management

International Nuclear Information System (INIS)

2005-03-01

Spent fuel management has become a prospective area for application of remote technology in recent years with a steadily growing inventory of spent fuel arising from nuclear power production. A remark that could be made from the review of technical information collected from the IAEA meetings was that remote technology in spent fuel management has matured well through the past decades of industrial experiences. Various remote technologies have been developed and applied in the past for facility operation and maintenance work in spent fuel examination, storage, transportation, reprocessing and radioactive waste treatment, among others, with significant accomplishments in dose reduction to workers, enhancement of reliability, etc. While some developmental activities are continuing for more advanced applications, industrial practices have made use of simple and robust designs for most of the remote systems technology applications to spent fuel management. In the current state of affairs, equipment and services in remote technology are available in the market for applications to most of the projects in spent fuel management. It can be concluded that the issue of critical importance in remote systems engineering is to make an optimal selection of technology and equipment that would best satisfy the as low as reasonably achievable (ALARA) requirements in terms of relevant criteria like dose reduction, reliability, costs, etc. In fact, good selection methodology is the key to efficient implementation of remote systems applications in the modern globalized market. This TECDOC gives a review of the current status of remote technology applications for spent fuel management, based on country reports from some Member States presented at the consultancy meetings, of which updated reports are attached in the annex. The scope of the review covers the series of spent fuel handling operations involved in spent fuel management, from discharge from reactor to reprocessing or
Burnup credit demands for spent fuel management in Ukraine

International Nuclear Information System (INIS)

Medun, V.

2001-01-01

In fact, till now, burnup credit has not be applied in Ukrainian nuclear power for spent fuel management systems (storage and transport). However, application of advanced fuel at VVER reactors, arising spent fuel amounts, represent burnup credit as an important resource to decrease spent fuel management costs. The paper describes spent fuel management status in Ukraine from viewpoint of subcriticality assurance under spent fuel storage and transport. It also considers: 1. Regulation basis concerning subcriticality assurance, 2. Basic spent fuel and transport casks characteristics, 3. Possibilities and demands for burnup credit application at spent fuel management systems in Ukraine. (author)
Spent fuel storage requirements 1993--2040

International Nuclear Information System (INIS)

1994-09-01

Historical inventories of spent fuel are combined with U.S. Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements through the year 2040. The needs are estimated for storage capacity beyond that presently available in the reactor storage pools. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of spent fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. The nuclear utilities provide historical data through December 1992 on the end of reactor life are based on the DOE/Energy Information Administration (EIA) estimates of future nuclear capacity, generation, and spent fuel discharges
Spent fuel shipping cask accident evaluation

International Nuclear Information System (INIS)

Fields, S.R.

1975-12-01

Mathematical models have been developed to simulate the dynamic behavior, following a hypothetical accident and fire, of typical casks designed for the rail shipment of spent fuel from nuclear reactors, and to determine the extent of radioactive releases under postulated conditions. The casks modeled were the IF-300, designed by the General Electric Company for the shipment of spent LWR fuel, and a cask designed by the Aerojet Manufacturing Company for the shipment of spent LMFBR fuel
Container for spent fuel assembly

International Nuclear Information System (INIS)

Sawai, Takeshi.

1996-01-01

The container of the present invention comprises a container main body having a body portion which can contain spent fuel assemblies and a lid, and heat pipes having an evaporation portion disposed along the outer surface of the spent fuel assemblies to be contained and a condensation portion exposed to the outside of the container main body. Further, the heat pipe is formed spirally at the evaporation portions so as to surround the outer circumference of the spent fuel assemblies, branched into a plurality of portions at the condensation portion, each of the branched portion of the condensation portion being exposed to the outside of the container main body, and is tightly in contact with the periphery of the slit portions disposed to the container main body. Then, since released after heat is transferred to the outside of the container main body from the evaporation portion of the heat pipe along the outer surface of the spent fuel assemblies by way of the condensation portion of the heat pipes exposed to the outside of the container main body, the efficiency of the heat transfer is extremely improved to enhance the effect of removing heat of spent fuel assemblies. Further, cooling effect is enhanced by the spiral form of the evaporation portion and the branched condensation portion. (N.H.)
Spent fuel critical masses and supportive measurements

International Nuclear Information System (INIS)

Toffer, H.; Wells, A.H.

1987-01-01

Critical masses for spent fuel are larger than for green fuel and therefore use of the increased masses could result in improved handling, storage, and transport of such materials. To apply spent fuel critical masses requires an assessment of fuel exposure and the corresponding isotopic compositions. The paper discusses several approaches at the Hanford N Reactor in establishing fuel exposure, including a direct measurement of spent to green fuel critical masses. The benefits derived from the use of spent fuel critical masses are illustrated for cask designs at the Nuclear Assurance Corporation. (author)
Granitoid emplacement during syn-convergent transtension: An example from the Huamenlou pluton in North Qinling, central China

Directory of Open Access Journals (Sweden)

Yang Li

2018-01-01

Full Text Available The Huamenlou pluton, is an elongated granite intrusion with high aspect ratio, emplaced within the southern margin of the North Qinling (central China. Here we investigate this pluton through multiple techniques including the fabric study, microstructural observation and zircon geochronology. Our zircon U–Pb data confirm that the granite crystallized at ca. 462 Ma which is consistent with the ages of other linear plutons in North Qinling. Microstructural observations of the Huamenlou granites illustrate that the pluton has undergone superimposed deformation during its emplacement, from magmatic to high-temperature solid state conditions. The internal fabric obtained by anisotropy of magnetic susceptibility (AMS and shape preferred orientation (SPO show similar results. The fabrics are relatively concordant and generally vary from NE–SW to NEE–SWW which are roughly oblique to the trend of the pluton elongation and the regional structures. Meanwhile, scalar parameters reflect two completely different strain regimes for the pluton and its host rocks, i.e., the fabrics within host rocks are mainly oblate while the central part of the intrusion displays mainly prolate fabrics. It is inferred that the structural pattern recorded in this pluton was caused by local dextral transtension in consequence of oblique convergence between the South and North China Blocks. We propose that the local transtension in convergence setting probably evolved from vertical extrusion tectonics that provided room for the magma emplacement and imparted prolate fabrics in the Huamenlou pluton.
Spent fuel pyroprocessing demonstration

International Nuclear Information System (INIS)

McFarlane, L.F.; Lineberry, M.J.

1995-01-01

A major element of the shutdown of the US liquid metal reactor development program is managing the sodium-bonded spent metallic fuel from the Experimental Breeder Reactor-II to meet US environmental laws. Argonne National Laboratory has refurbished and equipped an existing hot cell facility for treating the spent fuel by a high-temperature electrochemical process commonly called pyroprocessing. Four products will be produced for storage and disposal. Two high-level waste forms will be produced and qualified for disposal of the fission and activation products. Uranium and transuranium alloys will be produced for storage pending a decision by the US Department of Energy on the fate of its plutonium and enriched uranium. Together these activities will demonstrate a unique electrochemical treatment technology for spent nuclear fuel. This technology potentially has significant economic and technical advantages over either conventional reprocessing or direct disposal as a high-level waste option
Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program

Energy Technology Data Exchange (ETDEWEB)

Molecke, M.A.; Gregson, M.W.; Sorenson, K.B. [Sandia National Labs. (United States); Billone, M.C.; Tsai, H. [Argonne National Lab. (United States); Koch, W.; Nolte, O. [Fraunhofer Inst. fuer Toxikologie und Experimentelle Medizin (Germany); Pretzsch, G.; Lange, F. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (Germany); Autrusson, B.; Loiseau, O. [Inst. de Radioprotection et de Surete Nucleaire (France); Thompson, N.S.; Hibbs, R.S. [U.S. Dept. of Energy (United States); Young, F.I.; Mo, T. [U.S. Nuclear Regulatory Commission (United States)

2004-07-01

We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this research to better understand potential radiological impacts from sabotage of nuclear material shipments and storage casks, and to support subsequent risk assessments, modeling, and preventative measures. We provide a summary of the overall, multi-phase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on ''surrogate'' spent fuel, unirradiated depleted uranium oxide, and forthcoming actual spent fuel tests. The depleted uranium oxide test rodlets were prepared by the Institut de Radioprotection et de Surete Nucleaire, in France. These surrogate test rodlets closely match the diameter of the test rodlets of actual spent fuel from the H.B. Robinson reactor (high burnup PWR fuel) and the Surry reactor (lower, medium burnup PWR fuel), generated from U.S. reactors. The characterization of the spent fuels and fabrication into short, pressurized rodlets has been performed by Argonne National Laboratory, for testing at SNL. The ratio of the aerosol and respirable particles released from HEDD-impacted spent

Nondestructive verification and assay systems for spent fuels

International Nuclear Information System (INIS)

Cobb, D.D.; Phillips, J.R.; Bosler, G.E.; Eccleston, G.W.; Halbig, J.K.; Hatcher, C.R.; Hsue, S.T.

1982-04-01

This is an interim report of a study concerning the potential application of nondestructive measurements on irradiated light-water-reactor (LWR) fuels at spent-fuel storage facilities. It describes nondestructive measurement techniques and instruments that can provide useful data for more effective in-plant nuclear materials management, better safeguards and criticality safety, and more efficient storage of spent LWR fuel. In particular, several nondestructive measurement devices are already available so that utilities can implement new fuel-management and storage technologies for better use of existing spent-fuel storage capacity. The design of an engineered prototype in-plant spent-fuel measurement system is approx. 80% complete. This system would support improved spent-fuel storage and also efficient fissile recovery if spent-fuel reprocessing becomes a reality
An evaluation of the dissolution process of natural uranium ore as an analogue of nuclear fuel

International Nuclear Information System (INIS)

Stern, V.H.

1991-08-01

The assumption of congruent dissolution of uraninite as a mechanism for the dissolution behaviour of spent fuel was critically examined with regard to the fate of toxic radionuclides. The fission and daughter products of uranium are typically present in spent unreprocessed fuel rods in trace abundances. The principles of trace element geochemistry were applied in assessing the behaviour of these radionuclides during fluid/solid interactions. It is shown that the behaviour of radionuclides in trace abundances that reside in the crystal structure can be better predicted from the ionic properties of these nuclides rather than from assuming that they are controlled by the dissolution of uraninite. Geochemical evidence from natural uranium ore deposits (Athabasca Basin, Northern Territories of Australia, Oklo) suggests that in most cases the toxic radionuclides are released from uraninite in amounts that are independent of the solution behaviour of uranium oxide. Only those elements that have ionic and thus chemical properties similar to U 4+ , such as plutonium, americium, cadmium, neptunium and thorium can be satisfactorily modelled by the solution properties of uranium dioxide and then only if the environment is reducing. (84 refs., 7 tabs.)
Management of legacy spent nuclear fuel wastes at the Chalk River Laboratories: operating experience and progress towards waste remediation

International Nuclear Information System (INIS)

Cox, D.S.; Bainbridge, I.B.; Greenfield, K.R.

2006-01-01

AECL has been managing and storing a diversity of spent nuclear fuel, arising from operations at its Chalk River Laboratories (CRL) site over more than 50 years. A subset of about 22 tonnes of research reactor fuels, primarily metallic uranium, have been identified as a high priority for remediation, based on monitoring and inspection that has determined that these fuels and their storage containers are corroding. This paper describes the Fuel Packaging and Storage (FPS) project, which AECL has launched to retrieve these fuels from current storage, and to emplace them in a new above-ground dry storage system, as a prerequisite step to decommissioning some of the early-design waste storage structures at CRL. The retrieved fuels will be packaged in a new storage container, and subjected to a cold vacuum drying process that will remove moisture, and thereby reduce the extent of future corrosion and degradation. The FPS project will enable improved interim storage to be implemented for legacy fuels at CRL, until a decision is made on the ultimate disposition of legacy fuels in Canada. (author)
5 CFR 551.422 - Time spent traveling.

Science.gov (United States)

2010-01-01

... 5 Administrative Personnel 1 2010-01-01 2010-01-01 false Time spent traveling. 551.422 Section 551... Activities § 551.422 Time spent traveling. (a) Time spent traveling shall be considered hours of work if: (1... who is permitted to use an alternative mode of transportation, or an employee who travels at a time...
Spent fuel management newsletter. No. 2

International Nuclear Information System (INIS)

1993-04-01

This issue of the newsletter consists of two parts. The first part describes the IAEA Secretariat activities - work and programme of the Nuclear Materials and Fuel Cycle Technology Section of the Division of Nuclear Fuel Cycle and Waste Management, recent and planned meetings and publications, Technical Co-operation projects, Co-ordinated Research programmes. The second part contains country reports - national programmes on spent fuel management: current and planned storage and reprocessing capacities, spent fuel arisings, safety, transportation, storage and treatment of spent fuel
Spent fuel management newsletter. No. 2

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-04-01

This issue of the newsletter consists of two parts. The first part describes the IAEA Secretariat activities - work and programme of the Nuclear Materials and Fuel Cycle Technology Section of the Division of Nuclear Fuel Cycle and Waste Management, recent and planned meetings and publications, Technical Co-operation projects, Co-ordinated Research programmes. The second part contains country reports - national programmes on spent fuel management: current and planned storage and reprocessing capacities, spent fuel arisings, safety, transportation, storage and treatment of spent fuel.
Pyrochemical processing of DOE spent nuclear fuel

International Nuclear Information System (INIS)

Laidler, J.J.

1995-01-01

A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or open-quotes pyroprocessing,close quotes provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (>99.9%) separation of transuranics. The resultant waste forms from the pyroprocess, are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory
Lithofacies of deep marine basalts emplaced on a Jurassic backarc apron, Baja California (Mexico)

Energy Technology Data Exchange (ETDEWEB)

Busby-Spera, C.J.

1987-09-01

Basalts of the mid-Jurassic Gran Canon Formation, Cedros Island, Mexico, were emplaced on a volcaniclastic apron in a deep marine backarc basin. Elongate pillows and lava tubes, as well as paleocurrent data from the volcaniclastic apron, indicate a southward regional paleoslope away from the island arc source. Basalts emplaced on relatively proximal parts of the apron are nearly entirely pillowed and have thick flow units with mega-pillows. Basalts on distal parts of the apron (about 15 to 20 km down paleo-current) are dominated by pillow fragment breccias (flow foot rubble), and individual lava flows are generally thin, with small pillows, suggesting that the distal ends of lava flows, erupted upslope, are represented. These distal flow fronts, however, are interstratified with features that typically form close to a vent, including thick massive to mega-pillowed lavas and lava tubes up to 8 m in diameter. It is inferred that a fissure (or system of fissures) extended from the arc into the backarc basin, erupting basalt lavas onto both proximal and distal parts of the volcaniclastic apron. Such intraplate volcanism may be common on the hot frontal arc side of backarc basins. 26 references.
Characteristics of spent nuclear fuel

International Nuclear Information System (INIS)

Notz, K.J.

1988-04-01

The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the spent fuels and other wastes that will, or may, eventually be disposed of in a geological repository. The two major sources of these materials are commercial light-water reactor (LWR) spent fuel and immobilized high-level waste (HLW). Other wastes that may require long-term isolation include non-LWR spent fuels and miscellaneous sources such as activated metals. This report deals with spent fuels, but for completeness, the other sources are described briefly. Detailed characterizations are required for all of these potential repository wastes. These characteristics include physical, chemical, and radiological properties. The latter must take into account decay as a function of time. In addition, the present inventories and projected quantities of the various wastes are needed. This information has been assembled in a Characteristics Data Base which provides data in four formats: hard copy standard reports, menu-driven personal computer (PC) data bases, program-level PC data bases, and mainframe computer files. 5 refs., 3 figs., 4 tabs
National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel: construction accounting manual

International Nuclear Information System (INIS)

1978-12-01

The preparation of a cost estimate for a program to construct a new or add-on waste storage facility is one of the major objectives of the conceptual design effort. It is essential that a construction accounting manual be developed for use in estimating (conceptual, Title 1 and working estimates), cost accounting and cost control purposes on a consistent basis throughout life of the project. In order to meet this objective, itemized cost estimates and cost accounting data must be in sufficient detail to assure the reliability of such cost estimates and cost data. A cost accounting manual for contractors involved in the project is set forth in this document. For the conceptual design cost estimate, the estimate will be prepared in accordance with this construction accounting manual and all future estimates will be prepared on this basis. When work starts under the project, cost data will be accumulated and reported in accordance with this manual to provide consistency between estimates and actual cost data for cost control programs
Spent-fuel-storage alternatives

International Nuclear Information System (INIS)

1980-01-01

The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed
Granitic magma emplacement and deformation during early-orogenic syn-convergent transtension: The Stare Sedlo complex, Bohemian Massif

Czech Academy of Sciences Publication Activity Database

Tomek, Filip; Žák, J.; Chadima, Martin

2015-01-01

Roč. 87, JUL (2015), s. 50-66 ISSN 0264-3707 Institutional support: RVO:67985831 Keywords : anisotropy of magnetic susceptibility (AMS) * Bohemian Massif * pluton emplacement * granite * transtension * Variscan orogeny Subject RIV: DB - Geology ; Mineralogy Impact factor: 1.926, year: 2015
Spent fuel's behavior under dynamic drip tests

International Nuclear Information System (INIS)

Finn, P.A.; Buck, E.C.; Hoh, J.C.; Bates, J.K.

1995-01-01

In the potential repository at Yucca Mountain, failure of the waste package container and the cladding of the spent nuclear fuel would expose the fuel to water under oxidizing conditions. To simulate the release behavior of radionuclides from spent fuel, dynamic drip and vapor tests with spent nuclear fuel have been ongoing for 2.5 years. Rapid alteration of the spent fuel has been noted with concurrent release of radionuclides. Colloidal species containing americium and plutonium have been found in the leachate. This observation suggests that colloidal transport of radionuclides should be included in the performance assessment of a potential repository
Development of spent fuel dry storage technology

International Nuclear Information System (INIS)

Maruoka, Kunio; Matsunaga, Kenichi; Kunishima, Shigeru

2000-01-01

The spent fuels are the recycle fuel resources, and it is very important to store the spent fuels in safety. There are two types of the spent fuel interim storage system. One is wet storage system and another is dry storage system. In this study, the dry storage technology, dual purpose metal cask storage and canister storage, has been developed. For the dual purpose metal cask storage, boronated aluminum basket cell, rational cask body shape and shaping process have been developed, and new type dual purpose metal cask has been designed. For the canister storage, new type concrete cask and high density vault storage technology have been developed. The results of this study will be useful for the spent fuel interim storage. Safety and economical spent fuel interim storage will be realized in the near future. (author)
Thermal modeling of pluton emplacement and associated contact metamorphism:Parashi stock emplacement in the Serranía de Jarara (Alta Guajira, Colombia

Directory of Open Access Journals (Sweden)

Zuluaga C. Carlos A.

2010-12-01

Full Text Available
In the northernmost portion of the Serrania de Jarara (Alta Guajira, Colombia, low - medium grade metamorphic rocks from the Etpana Metamorphic Suite were thermally affected by emplacement of a small calc-alkaline intrusion (Parashi Stock. Detailed petrographic analysis in collected rock samples across the NE and NW plutonic contacts show occurrences of textural and mineralogical changes in the country rock fabric that evidence contact metamorphism overprinting regional metamorphism of the Etpana Suite. These changes include growth of andalusite (chiastolite, calcic clinopyroxeneand amphibole porphyroblast crosscutting Sn+1 metamorphicfoliation. Hornblende-plagioclase barometry (ca. 3.1 kbar and cooling models for the stock show maximum time temperature evolution in the country rock at the interpreted depth of intrusion (ca. 11 km and help to evaluate the behavior of the country rock with the changing local geotherm.
Intermodal transfer of spent fuel

International Nuclear Information System (INIS)

Neuhauser, K.S.; Weiner, R.F.

1991-01-01

As a result of the international standardization of containerized cargo handling in ports around the world, maritime shipment handling is particularly uniform. Thus, handier exposure parameters will be relatively constant for ship-truck and ship-rail transfers at ports throughout the world. Inspectors' doses are expected to vary because of jurisdictional considerations. The results of this study should be applicable to truck-to-rail transfers. A study of the movement of spent fuel casks through ports, including the loading and unloading of containers from cargo vessels, afforded an opportunity to estimate the radiation doses to those individuals handling the spent fuels with doses to the public along subsequent transportation routes of the fuel. A number of states require redundant inspections and for escorts over long distances on highways; thus handlers, inspectors, escort personnel, and others who are not normally classified as radiation workers may sustain doses high enough to warrant concern about occupational safety. This paper addresses the question of radiation safety for these workers. Data were obtained during, observation of the offloading of reactor spent fuel (research reactor spent fuel, in this instance) which included estimates of exposure times and distances for handlers, inspectors and other workers during offloading and overnight storage. Exposure times and distance were also for other workers, including crane operators, scale operators, security personnel and truck drivers. RADTRAN calculational models and parameter values then facilitated estimation of the dose to workers during incident-free ship-to-truck transfer of spent fuel
Spent fuel dissolution mechanisms

International Nuclear Information System (INIS)

Ollila, K.

1993-11-01

This study is a literature survey on the dissolution mechanisms of spent fuel under disposal conditions. First, the effects of radiolysis products on the oxidative dissolution mechanisms and rates of UO 2 are discussed. These effects have mainly been investigated by using electrochemical methods. Then the release mechanisms of soluble radionuclides and the dissolution of the UO 2 matrix including the actinides, are treated. Experimental methods have been developed for measuring the grain-boundary inventories of radionuclides. The behaviour of cesium, strontium and technetium in leaching tests shows different trends. Comparison of spent fuel leaching data strongly suggests that the release of 90 Sr into the leachant can be used as a measure of the oxidation/dissolution of the fuel matrix. Approaches to the modelling UO 2 , dissolution are briefly discussed in the next chapter. Lastly, the use of natural material, uraninite, in the evaluation of the long-term performance of spent fuel is discussed. (orig.). (81 ref., 37 figs., 8 tabs.)
Storing the world's spent nuclear fuel

International Nuclear Information System (INIS)

Barkenbus, J.N.; Weinberg, A.M.; Alonso, M.

1985-01-01

Given the world's prodigious future energy requirements and the inevitable depletion of oil and gas, it would be foolhardy consciously to seek limitations on the growth of nuclear power. Indeed, the authors continue to believe that the global nuclear power enterprise, as measured by installed reactor capacity, can become much larger in the future without increasing proliferation risks. To accomplish this objective will require renewed dedication to the non-proliferation regime, and it will require some new initiatives. Foremost among these would be the establishment of a spent fuel take-back service, in which one or a few states would retrieve spent nuclear fuel from nations generating it. The centralized retrieval of spent fuel would remove accessible plutonium from the control of national leaders in non-nuclear-weapons states, thereby eliminating the temptation to use this material for weapons. The Soviets already implement a retrieval policy with the spent fuel generated by East European allies. The authors believe that it is time for the US to reopen the issue of spent-fuel retrieval, and thus to strengthen its non-proliferation policies and the nonproliferation regime in general. 7 references
Subduction initiation, recycling of Alboran lower crust, and intracrustal emplacement of subcontinental lithospheric mantle in the Westernmost Mediterranean

Science.gov (United States)

Varas-Reus, María Isabel; Garrido, Carlos J.; Bosch, Delphine; Marchesi, Claudio; Hidas, Károly; Booth-Rea, Guillermo; Acosta-Vigil, Antonio

2015-04-01

Unraveling the tectonic settings and processes involved in the annihilation of subcontinental mantle lithosphere is of paramount importance for our understanding of the endurance of continents through Earth history. Unlike ophiolites -- their oceanic mantle lithosphere counterparts -- the mechanisms of emplacement of the subcontinental mantle lithosphere in orogens is still poorly known. The emplacement of subcontinental lithospheric mantle peridotites is often attributed to extension in rifted passive margins or continental backarc basins, accretionary processes in subduction zones, or some combination of these processes. One of the most prominent features of the westernmost Mediterranean Alpine orogenic arcs is the presence of the largest outcrops worldwide of diamond facies, subcontinental mantle peridotite massifs; unveiling the mechanisms of emplacement of these massifs may provide important clues on processes involved in the destruction of continents. The western Mediterranean underwent a complex Alpine evolution of subduction initiation, slab fragmentation, and rollback within a context of slow convergence of Africa and Europe In the westernmost Mediterranean, the alpine orogeny ends in the Gibraltar tight arc, which is bounded by the Betic, Rif and Tell belts that surround the Alboran and Algero-Balearic basins. The internal units of these belts are mostly constituted of an allochthonous lithospheric domain that collided and overthrusted Mesozoic and Tertiary sedimentary rocks of the Mesozoic-Paleogene, South Iberian and Maghrebian rifted continental paleomargins. Subcontinental lithospheric peridotite massifs are intercalated between polymetamorphic internal units of the Betic (Ronda, Ojen and Carratraca massifs), Rif (Beni Bousera), and Tell belts. In the Betic chain, the internal zones of the allochthonous Alboran domain include, from bottom to top, polymetamorphic rock of the Alpujarride and Malaguide complexes. The Ronda peridotite massif -- the
Spent-fuel-storage alternatives

Energy Technology Data Exchange (ETDEWEB)

1980-01-01

The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

SOCIAL INTEGRATION: TESTING ANTECEDENTS OF TIME SPENT ONLINE

Directory of Open Access Journals (Sweden)

Lily Suriani Mohd Arif

2013-07-01

Full Text Available The literature on the relationship of social integration and time spent onlineprovides conflicting evidence of the relationship of social integration with timespent online. The study identifies and highlightsthe controversy and attempts toclarify the relationship of social integration withtime spent online bydecomposing the construct social integration into its affective and behavioraldimensions . Thestudy tests antecedents and effects of time spent online in arandom sample of senior level undergraduate students at a public university inMalaysia. The findings indicated that while self-report measures of behavioralsocial integration did not predict time spent online, and, the affective socialintegration had an inverse relationship with time spent online.
Pluton emplacement in a releasing bend in a transpressive regime: the arrozal granite in the Paraíba do Sul shear belt, Rio de Janeiro.

Science.gov (United States)

Nummer, Alexis R; Machado, Rômulo; Dehler, Nolan M

2007-06-01

The Arrozal Granite, situated in the southwestern region of the State of Rio de Janeiro, has a granitic to granodioritic composition. It contains a strong mylonitic foliation along its border, passing gradually to a well-developed magmatic foliation towards its center. Structural analysis indicates that the Arrozal Granite was emplaced along the Além-Paraíba Shear Zone in a dextral transpressive tectonic regime. A regional shift of the trend along this shear zone from NE-SW to E-W, observed in the area, is interpreted to be casually related to the creation of space for the emplacement of the granite. Our data indicate that releasing bends may have played an important role for space generation during the emplacement of the Arrozal Granite and other plutons.
Spent Pot Lining Characterization Framework

Science.gov (United States)

Ospina, Gustavo; Hassan, Mohamed I.

2017-09-01

Spent pot lining (SPL) management represents a major concern for aluminum smelters. There are two key elements for spent pot lining management: recycling and safe storage. Spent pot lining waste can potentially have beneficial uses in co-firing in cement plants. Also, safe storage of SPL is of utmost importance. Gas generation of SPL reaction with water and ignition sensitivity must be studied. However, determining the feasibility of SPL co-firing and developing the required procedures for safe storage rely on determining experimentally all the necessary SPL properties along with the appropriate test methods, recognized by emissions standards and fire safety design codes. The applicable regulations and relevant SPL properties for this purpose are presented along with the corresponding test methods.
Spent fuel storage requirements, 1988

International Nuclear Information System (INIS)

1988-10-01

Historical inventories of spent fuel and Department of Energy (DOE) estimates of future discharges from US commercial nuclear reactors are presented for the next 20 years, through the year 2007. The eventual needs for additional spent fuel storage capacity are estimated. These estimates are based on the maximum capacities within current and planned at-reactor facilities and on any planned transshipments of fuel to other reactors or facilities. Historical data through December 1987 and projected discharges through the end of reactor life are used in this analysis. The source data was supplied by the utilities to DOE through the 1988 RW-859 data survey and by DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 12 refs., 3 figs., 28 tabs
Thermal model of spent fuel transport cask

International Nuclear Information System (INIS)

Ahmed, E.E.M.; Rahman, F.A.; Sultan, G.F.; Khalil, E.E.

1996-01-01

The investigation provides a theoretical model to represent the thermal behaviour of the spent fuel elements when transported in a dry shipping cask under normal transport conditions. The heat transfer process in the spent fuel elements and within the cask are modeled which include the radiant heat transfer within the cask and the heat transfer by thermal conduction within the spent fuel element. The model considers the net radiant method for radiant heat transfer process from the inner most heated element to the surrounding spent elements. The heat conduction through fuel interior, fuel-clad interface and on clad surface are also presented. (author) 6 figs., 9 refs
The Summer 1997 Eruption at Pillan Patera on Io: Implications for Ultrabasic Lava Flow Emplacement

Science.gov (United States)

Williams, David A.; Davies, Ashley G.; Keszthelyi, Laszlo P.; Greeley, Ronald

2001-01-01

Galileo data and numerical modeling were used to investigate the summer 1977 eruption at Pillan Patera on Io. This event, now defined as "Pillanian" eruption style, included a high-temperature (greater than 1600 C), possible ultrabasic , 140-km-high plume eruption that deposited dark, orthopyroxene-rich pyroclastic material over greater than 125,000 sq km, followed by emplacement of dark flow-like material over greater than 3100 sq km to the north of the caldera. We estimate that the high-temperature, energetic episode of this eruption had a duration of 52 - 167 days between May and September 1997, with peak eruption temperatures around June 28, 1997. Galileo 20 m/pixel images of part of the Pillan flow field show a wide-spread, rough, pitted surface that is unlike any flow surface we have seen before. We suggest that this surface may have resulted from: 1. A fractured lava crust formed during rapid, low-viscosity lava surging, perhaps including turbulent flow emplacement. 2. Disruption of the lava flow by explosive interaction with a volatile-rich substrate. or 3. A combination of 1 and 2 with or without accumulation of pyroclastic material on the surface. Well-developed flow lobes are observed, suggesting that this is a relatively distant part of the flow field.Shadow measurements at flow margins indicate a thickness of-8 - 10 m. We have modeled the emplacement of putative ultrabasic flow from the summer 1997 Pillan eruption using constraints from new Galileo data. Results suggest that either laminar sheet flows or turbulent channelized flows could have traveled 50 - 150 km on a flat, unobstructed surface, which is consistent with the estimated length of the Pillan flow field (approx. 60 km). Our modeling suggests low thermal erosion rates (less than 4.1 m/d), and that the formation of deep (greater than 20 m) erosion channels was unlikely, especially distal to the source. We calculate a volumetric flow rate of approx. 2 - 7 x 10(exp 3)cu m/s, which is greater
Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio programme

International Nuclear Information System (INIS)

Molecke, M.A.; Gregson, M.W.; Sorenson, K.B.

2004-01-01

We provide a detailed overview of an on-going, multinational test programme that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolised materials plus volatilised fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy/density device. The programme participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research programme. Sandia National Laboratories has the lead role for conducting this research programme; test programme support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. We provide a summary of the overall, multiphase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on 'surrogate' spent fuel, unirradiated depleted uranium oxide and forthcoming actual spent fuel tests. We briefly summarise similar results from completed surrogate tests that used non-radioactive, sintered cerium oxide ceramic pellets in test rods. (author)
Spent fuel management in France: Programme status

International Nuclear Information System (INIS)

Chaudat, J.P.

1990-01-01

France's programme is best characterized as a closed fuel cycle including reprocessing, Plutonium recycling in PWR and use of breeder reactors. The current installed nuclear capacity is 52.5 GWe from 55 units. The spent fuel management scheme chosen is reprocessing. This paper describes the national programme, spent nuclear fuel storage, reprocessing and contracts for reprocessing of spent fuel from various countries. (author). 5 figs, 2 tabs
Fabrication of the Spent Fuel Elements Rack on the ISFSF

International Nuclear Information System (INIS)

Slamet Wiranto; Sigit Purwanto; Safrul, H.

2004-01-01

The Interim Storage For Spent Fuel elements (ISFSF) was designed to be able to store the 33 spent fuel element racks with capacity of 1386 of normal spent fuel elements and 2 racks for 36 of defected ones. Until now, only 9 out of 33 racks of normal spent fuel elements and lout of 2 racks of defected fuel elements are available. Five of them have suffered from corrosion so that they are not fulfilled the requirements of the spent fuel elements storage anymore. Meanwhile, the spent fuel storage racks in the reactor are almost full. It means, the transfer of the spent fuel from reactor spent fuel storage to the ISFSF pool are compulsory needed. Therefore, it is necessary to provide the new ISFSF spent fuel storage rack with better material and fabrication method than the old one. In this design all materials consist of SS 316 L that are welded with the Argon TIG-welding. Right now there has been one new spent fuel storage rack fabricated with capacity of 42 normal spent fuel elements. (author)
Storage of Spent Nuclear Fuel. Specific Safety Guide

International Nuclear Information System (INIS)

2012-01-01

This Safety Guide provides recommendations and guidance on the storage of spent nuclear fuel. It covers all types of storage facilities and all types of spent fuel from nuclear power plants and research reactors. It takes into consideration the longer storage periods that have become necessary owing to delays in the development of disposal facilities and the decrease in reprocessing activities. It also considers developments associated with nuclear fuel, such as higher enrichment, mixed oxide fuels and higher burnup. The Safety Guide is not intended to cover the storage of spent fuel if this is part of the operation of a nuclear power plant or spent fuel reprocessing facility. Guidance is provided on all stages for spent fuel storage facilities, from planning through siting and design to operation and decommissioning, and in particular retrieval of spent fuel. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. Management system; 5. Safety case and safety assessment; 6. General safety considerations for storage of spent fuel. Appendix I: Specific safety considerations for wet or dry storage of spent fuel; Appendix II: Conditions for specific types of fuel and additional considerations; Annex: I: Short term and long term storage; Annex II: Operational and safety considerations for wet and dry spent fuel storage facilities; Annex III: Examples of sections of operating procedures for a spent fuel storage facility; Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex VI: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex VII: Postulated initiating events for consideration in a safety assessment (internal phenomena).
A study on the safety of spent fuel management. A scenario study on spent fuel management

Energy Technology Data Exchange (ETDEWEB)

Chun, Kwan Sik; Park, Hyun Soo; Ahn, Jin Soo; Hwang, Joo Ho; Choi, Jong Won; Kim, Yeon Soo; Park, Ju Hwan; Chung, Choong Hwan [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

1992-03-01

In order to produce data applicable for the long-term policy making of spent fuel management and to suggest a basic scenario suitable to domestic situation, the pre-conceptual design of reference disposal facilities for the spent fuel and the vitrified high level radioactive waste from its reprocessing, has been performed. From the results of the pre-conceptual study, further research and development areas to accumulate the disposal technology are suggested. In addition, the physico-chemical properties and functional characteristics of domestic bentonite are analyzed to assess its applicability as a buffer material which would play a major role for the safe disposal of highly active waste including spent fuels. (Author).
Radioactive waste management decommissioning spent fuel storage. V. 3. Waste transport, handling and disposal spent fuel storage

International Nuclear Information System (INIS)

1985-01-01

As part of the book entitled Radioactive waste management decommissioning spent fuel storage, vol. 3 dealts with waste transport, handling and disposal, spent fuel storage. Twelve articles are presented concerning the industrial aspects of nuclear waste management in France [fr
Waste package transfer, emplacement and retrievability in the French deep geological repository

Energy Technology Data Exchange (ETDEWEB)

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

2009-06-15

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

International Nuclear Information System (INIS)

Gerstner, D.M.; Clayton, S.G.; Farrell, R.F.; McCormick, J.A.; Ortiz, C.; Standiford, D.L.

1996-01-01

The Carlsbad Area Office established and analyzed the safety bases for the design and operations as documented in the WIPP Safety Analysis Report (SAR). Additional independent efforts are currently underway to assess the hazards associated with the long-term (10,000 year) isolation period as required by 40 CFR 191. The structure of the WIPP SAR is unique due to the hazards involved, and the agreement between the State of New Mexico and the DOE regarding SAR content and format. However, the hazards and accident analysis philosophy as contained in DOE-STD-3009-94 was followed as closely as possible, while adhering to state agreements. Hazards associated with WIPP waste receipt, emplacement, and disposal operations were systematically identified using a modified Hazard and Operability Study (HAZOP) technique. The WIPP HAZOP assessed the potential internal, external, and natural phenomena events that can cause the identified hazards to develop into accidents. The hazard assessment identified deviations from the intended design and operation of the waste handling system, analyzed potential accident consequences to the public and workers, estimated likelihood of occurrence, and evaluated associated preventative and mitigative features. It was concluded from the assessment that the proposed WIPP waste emplacement operations and design are sufficient to ensure safety of the public, workers, and environment, over the 35 year disposal phase
Emplacement dynamics and lava field evolution of the flood basalt eruption at Holuhraun, Iceland: Observations from field and remote sensing data

Science.gov (United States)

Pedersen, Gro; Höskuldsson, Armann; Riishuus, Morten S.; Jónsdóttir, Ingibjörg; Thórdarson, Thorvaldur; Dürig, Tobias; Gudmundsson, Magnus T.; Durmont, Stephanie

2016-04-01

The Holuhraun eruption (Aug 2014- Feb 2015) is the largest effusive eruption in Iceland since the Laki eruption in 1783-84, with an estimated lava volume of ~1.6 km3 covering an area of ~83 km2. The eruption provides an unprecedented opportunity to study i) lava morphologies and their emplacement styles, ii) Morphological transitions iii) the transition from open to closed lava pathways and iv) the implication of lava pond formation. This study is based on three different categories of data; field data, airborne data and satellite data. The field data include tracking of the lava advancement by Global Positioning System (GPS) measurements and georeferenced GoPro cameras allowing classification of the lava margin morphology. Furthermore, video footage on-site documented lava emplacement. Complimentary observations have been provided from aircraft platforms and by satellite data. Of particular importance for lava morphology observations are 1-12 m/pixel airborne Synthetic Aperture Radar (SAR) images (x-band), as well as SAR data from TerraSAR-X and COSMO-SkyMed satellites. The Holuhraun lava field comprises a continuum of morphologies from pāhoehoe to 'a'ā, which have varied temporally and spatially. Shelly pāhoehoe lava was the first morphology to be observed (08-29). Spatially, this lava type was not widely distributed, but was emplaced throughout the eruption close to the vent area and the lava channels. Slabby pāhoehoe lava was initially observed the 08-31 and was observed throughout most of the eruption during the high-lava-flux phase of new lava lobe emplacement. 'A'ā lavas were the dominating morphology the first three months of the eruption and was first observed 09-01 like Rubbly pāhoehoe lava. Finally, Spiny pāhoehoe lava was first observed the 09-05 as a few marginal outbreaks along the fairly inactive parts of the 'a'ā lava lobe. However, throughout the eruption this morphology became more important and from mid-November/beginning of December the
Spent fuel management: Current status and prospects 1993

International Nuclear Information System (INIS)

1994-02-01

Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and it is still one of the most vital problems common to all countries with nuclear reactors. It begins with the discharge of spent fuel from a power or a research reactor and ends with its ultimate disposition, either by direct disposal or by reprocessing of the spent fuel. Two options exist at present - an open, once-through cycle with direct disposal of the spent fuel and a closed cycle with reprocessing of the spent fuel and recycling of plutonium and uranium in new mixed oxide fuels. The selection of a spent fuel strategy is a complex procedure in which many factors have to be weighed, including political, economic and safeguards issues as well as protection of the environment. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for the exchange of information and to co-ordinate and to encourage closer co-operation among Member States in certain research an development activities that are of common interest. Refs, figs and tabs
Past and future IAEA spent fuel management activities

International Nuclear Information System (INIS)

Grigoriev, A.

1993-01-01

The main objectives and strategies of the Agency's activities in the area of spent fuel management are to promote the exchange of information between Member States on technical, safety, environmental and economic aspects of spent fuel management technology, including storage, transport and treatment of spent fuel, and to provide assistance to Member States in the planning, implementation and operation of nuclear fuel cycle facilities. This paper give a list of the meetings held since the last issue of the Spent Fuel Management Newsletter
Characterization of monolith block of spent resin cementation

International Nuclear Information System (INIS)

Prayitno; Endro-Kismolo; Isman MT

1996-01-01

Spent resin immobilization process with cement was done to prevent release of radionuclide in the ultimate storage or disposal. The varied Composition of water/cement ratio in the cementation process were 0.3; 0.4; 0.5 and the various weight of resin waste are 25 g, 37.5 g and 50 gram. The compressive strength of the various water/cement ratio without spent resin was bigger than 0.3. This investigation proved that the compressive strength of Tiga Roda cement was bigger than those of Gresik cement or Nusantara cement. The compressive of the cement block of were the spent resin cementation was influenced by the water/cement ratio and the total spent resin addition. The best condition reached at the water/cement ratio of 0.3 and 25 gram spent resin, was compressive strength of 17.86 N/mm 2 . Leaching rate of the various weight composition of spent resin cementation for 91 days were between 10 -2 - 10 -4 gram.cm -2 .day -1
Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

Energy Technology Data Exchange (ETDEWEB)

Ham, Y.S.; Kerr, P.; Sitaraman, S.; Swan, R. [Global Security Directorate, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Rossa, R. [SCK-CEN, Mol (Belgium); Liljenfeldt, H. [SKB in Oskarshamn (Sweden)

2015-07-01

The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)
A review of the ascent and emplacement of granitoid bodies into the crust

Directory of Open Access Journals (Sweden)

Katarína Bónová

2005-03-01

Full Text Available This paper relates to basic information (i.e. mechanical aspects of ascent, indicators faciliting the discriminability of various ascent styles about the models of ascent and emplacement of granitoid bodies, since the purely mechanical aspect of intrusion of magmas is a fascinating subject and it has generated a considerable controversy over many years. Individual models are demonstrated by world-known occurrences and examples from Western Carpathians region. The conditions of magma migration are demonstrated as well.

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Evolution report

International Nuclear Information System (INIS)

Smith, P.; Johnson, L.; Snellman, M.; Pastina, B.; Gribi, P.

2007-12-01

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 method. Posiva and SKB have conducted a joint research, demonstration and development (RD and D) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007, have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is pre-packaged in a perforated steel cylinder prior to emplacement in the deposition drift; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100 - 300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the repository evolution in successive time frames, including key uncertainties. The description of evolution starts with the initial conditions at the time of emplacement of the first canisters. The repository evolves through an early, transient phase to a state where evolution is far slower. Particular attention is given to describing the transient phase, since this is where most of the
Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Evolution report

International Nuclear Information System (INIS)

Smith, Paul; Johnson, Lawrence; Snellman, Margit; Pastina, Barbara; Gribi, Peter

2008-01-01

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007, have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is pre-packaged in a perforated steel cylinder prior to emplacement in the deposition drift; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the repository evolution in successive time frames, including key uncertainties. The description of evolution starts with the initial conditions at the time of emplacement of the first canisters. The repository evolves through an early, transient phase to a state where evolution is far slower. Particular attention is given to describing the transient phase, since this is where most of the
Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Evolution report

Energy Technology Data Exchange (ETDEWEB)

Smith, Paul; Johnson, Lawrence; Snellman, Margit; Pastina, Barbara; Gribi, Peter

2008-01-15

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007, have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is pre-packaged in a perforated steel cylinder prior to emplacement in the deposition drift; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the repository evolution in successive time frames, including key uncertainties. The description of evolution starts with the initial conditions at the time of emplacement of the first canisters. The repository evolves through an early, transient phase to a state where evolution is far slower. Particular attention is given to describing the transient phase, since this is where most of the
Research reactor spent fuel in Ukraine

International Nuclear Information System (INIS)

Trofimenko, A.P.

1996-01-01

This paper describes the research reactors in Ukraine, their spent fuel facilities and spent fuel management problems. Nuclear sciences, technology and industry are highly developed in Ukraine. There are 5 NPPs in the country with 14 operating reactors which have total power capacity of 12,800 MW
Modeling the highway transportation of spent fuel

International Nuclear Information System (INIS)

Harrison, I.G.

1986-01-01

There will be a substantial increase in the number of spent fuel shipments on the nation's highway system in the next thirty years. Most of the spent fuel will be moving from reactors to a spent fuel repository. This study develops two models that evaluate the risk and cost of moving the spent fuel. The Minimum Total Transport Risk Model (MTTRM) seeks an efficient solution for this problem by finding the minimum risk path through the network and sending all the spent fuel shipments over this one path. The Equilibrium Transport Risk Model (ETRM) finds an equitable solution by distributing the shipments over a number of paths in the network. This model decreases the risk along individual paths, but increases society's risk because the spent fuel shipments are traveling over more links in the network. The study finds that there is a trade off between path risk and societal risk. As path risk declines, societal risk rises. The cost of shipping also increases as the number of paths expand. The cost and risk of shipping spent fuel from ten reactors to four potential repository sites are evaluated using the MTTRM. The temporary monitored retrievable storage (MRS) facility in Tennessee is found to be the minimum cost and minimum risk solution. When direct shipment to the permanent sites is considered, Deaf Smith, Texas is the least cost and least incident free transport risk location. Yucca Mountain, Nevada is the least risk location when the focus is placed on the potential consequences of an accident
EVALUATION OF RISKS AND WASTE CHARACTERIZATION REQUIREMENTS FOR THE TRANSURANIC WASTE EMPLACED IN WIPP DURING 1999

International Nuclear Information System (INIS)

Channell, J.K.; Walker, B.A.

2000-01-01

Specifically this report: 1. Compares requirements of the WAP that are pertinent from a technical viewpoint with the WIPP pre-Permit waste characterization program, 2. Presents the results of a risk analysis of the currently emplaced wastes. Expected and bounding risks from routine operations and possible accidents are evaluated; and 3. Provides conclusions and recommendations
Basalt alteration and basalt-waste interaction in the Pasco Basin of Washington State. Final report

International Nuclear Information System (INIS)

Benson, L.V.; Carnahan, C.L.; Apps, J.A.; Mouton, C.A.; Corrigan, D.J.; Frisch, C.J.; Shomura, L.K.

1978-09-01

A study was conducted to determine the nature of the minerals which coat vesicle and fracture surfaces in the Grande Ronde Basalt Formation, simulate the mass transfer which led to their precipitation, and predict the mass transfer associated with the dissolution of spent unreprocessed fuel (SURF). Scanning electron microscopy (SEM), petrographic, x-ray diffraction (XRD), and electron microprobe (EMP) analyses have been made on a series of samples taken from 1100 ft (335.3 m) of core from core hole DC2. Preliminary simulations of the mass transfer associated with basalt dissolution in a thermodynamically closed system have been accomplished. In addition two mass transfer codes have been modified to facilitate data base changes. Thermochemical data for uranium and plutonium have been collected and converted to standard state conditions. These data will be critically evaluated and input to the mass transfer data base in the near future
Contribution to draft generic environmental impact statement on commercial waste management: radioactive waste isolation in geologic formations

International Nuclear Information System (INIS)

1978-04-01

This document concentrates on deep geologic isolation of wastes in bedded salt, granite, shale, and basalt with emphasis on wastes from three fuel cycles: reprocessing wastes from uranium and plutonium recycling, reprocessing wastes from uranium-only recycling, and spent unreprocessed fuel with no recycling. The analyses presented in this document are based on preconceptual repository designs. As the repository designs progress through future phases, refinements will occur which might modify some of these results. The 12 sections in the report are: introduction; selection and description of generic repository sites; LWR wastes to be isolated in geologic formations; description of waste isolation facilities; effluents from the waste isolation facility; assessment of environment impacts for various geographical locations of a waste isolation facility; environmental monitoring; decommissioning; mine decommissioning site restoration; deep geologic alternative actions; potential mechanisms of containment failure; and considerations relevant to provisional versus final storage
Spent fuel management in South Africa

International Nuclear Information System (INIS)

Bredell, P.J.; Stott, A.K.

1998-01-01

Eskom, the South African utility, operates one of the largest electricity networks in the world. However, only 6% of the South African generating capacity is nuclear; the remainder is coal fired and hydroelectric. The nuclear component consists of the Koeberg Nuclear Power Plant, comprising two French supplied PWRs of 920 MWe each, situated approximately 45 kilometres from cape Town. Construction started in 1976 and the two reactors reached criticality in 1984 and 1985 respectively. South Africa also has an Oak Ridge type research reactor, called SAFARI, operated by the South African Atomic Energy Corporation (AEC) at their Pelindaba site near Pretoria. This research reactor was commissioned in 1965, and has been in operation ever since. South Africa has a National Radioactive Waste Disposal facility called Vaalputs, some 600 km north of Cape Town. The facility, operated by AEC, is presently licensed only for the disposal of low and intermediate radioactive level wastes. Vaalputs offers unique features as a potential interim spent fuel storage and final disposal site, such as favorable geology (granite), low seismicity, low population density, remoteness from industrial centres and and conditions. Therefore, this site has been investigated by the AEC as a potential interim spent fuel storage site, but has not yet been licensed for this purpose. Hence, all spent fuel is currently stored on the two sites at Koeberg and Pelindaba respectively. The spent fuel storage pools at Koeberg have recently been enlarged to accommodate the lifetime spent fuel arisings of the plant. Since late 1997, the Safari spent fuel is stored in a pipe storage facility, constructed away from the reactor on the Pelindaba site. (author)
Status of research reactor spent fuel world-wide

International Nuclear Information System (INIS)

Ritchie, I.G.

2004-01-01

Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialised and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author)
Management of Spent Radiation Source from Radiotherapy

International Nuclear Information System (INIS)

Aisyah

2008-01-01

Nowadays the use of radioactive source for both radiodiagnostic and radiotherapy in Indonesia hospital increases rapidly. Sealed source used in radiotherapy among others for brachytherapy, teletherapy, bone densitometry, whole blood irradiation and gamma knife (radiosurgery). In line with this, the waste of spent radiation sources will be generated in hospitals. Of course these spent radiation sources must be treated correctly in order to maintain the safety of both the public and the environment. According to the Act No. 10/1997, BATAN, in care of the Radioactive Waste Management Center is the national appointed agency for the management of radioactive waste. The option for waste management by hospitals needs to be expound, either by re-exporting to the supplier of origin, re-exporting to other supplier, re-use by other licensee or sending to the Radioactive Waste Management Center. Usually the waste sent by the hospitals to the center comprises of sealed radiation source of 60 Co, 137 Cs or 226 Ra. The management of spent radiation source in the center is carried out in several steps i.e. conditioning, temporary storage, and long-term storage (final disposal). The conditioning of non 226 Ra is carried out by placing the waste in a 200 litter drum shell, 950 or 350 litter concrete shells, depends on the activity and dimension of the spent radiation source. The conditioning of 226 Ra is carried out by encapsulating the waste in a stainless steel container for long-term storage shield which then placed in a 200 litter drum shell. The temporary storage of the conditioned spent radiation source is carried out by storing it in the center’s temporary storages, either low or medium activity waste. Finally, the conditioned spent radiation source is buried in a disposal facility. For medium half-life spent radiation source, the final disposal is burial it in a shallow-land disposal; mean while, for long half-life spent radiation source, the final disposal is burial it in
Spent nuclear fuel transport problems

International Nuclear Information System (INIS)

Kondrat'ev, A.N.; Kosarev, Yu.A.; Yulikov, E.I.

1977-01-01

The paper considers the problems of shipping spent fuel from nuclear power stations to reprocessing plants and also the principal ways of solving these problems with a view to achieving maximum economy and safety in transport. The increase in the number of nuclear power plants in the USSR will entail an intensification of spent-fuel shipments. Higher burnup and the need to reduce cooling time call for heavier and more complex shipping containers. The problem of shipping spent fuel should be tackled comprehensively, bearing in mind the requirements of safety and economy. One solution to these problems is to develop rational and cheap designs of such containers. In addition, the world-wide trend towards more thorough protection of the environment against pollution and of the health of the population requires the devotion of constant attention to improving the reliability and safety of shipments. The paper considers the prospects for nuclear power development in the USSR and in other member countries of the CMEA (1976-1980), the composition and design of some Soviet packaging assemblies, the appropriate cooling time for spent fuel from thermal reactor power stations, procedures for reducing fuel-shipping costs, some methodological problems of container calculation and design, and finally problems of testing and checking containers on test rigs. (author)
Reracking Possibilities of the NPP Krsko Spent Fuel Pool

International Nuclear Information System (INIS)

Bace, M.; Pevec, D.; Smuc, T.

1998-01-01

Using the SCALE-4 code package reracking possibilities of the NPP Krsko spent fuel pool were analyzed. Two cases were considered: the first case assuming the 40 years lifetime of the plant, and the second case assuming the 50 years lifetime of the plant. It was shown that it is possible to design the additional racks in free space of the spent fuel pool with the sufficient total capacity to store all the spent fuel generated during the 40 years lifetime of the plant. In the case of 50 years plant lifetime, completely new racks (capacity of 1890 spent fuel assemblies), containing 4mm boral in storage cell walls, were proposed for the NPP Krsko spent fuel pool. The effective multiplication factor of the spent fuel pool fully loaded with new racks containing spent fuel assemblies of initial enrichment 4.3 w/o, burned to 40 GWd/tU and cooled 2 years is lower than the value required by standard. It showed the possibility of the safe disposal of all spent fuel accumulated during more than 50 years lifetime of the plant. (author)
Development of advanced spent fuel management process

International Nuclear Information System (INIS)

Park, Seong Won; Shin, Y. J.; Cho, S. H.

2004-03-01

The research on spent fuel management focuses on the maximization of the disposal efficiency by a volume reduction, the improvement of the environmental friendliness by the partitioning and transmutation of the long lived nuclides, and the recycling of the spent fuel for an efficient utilization of the uranium source. In the second phase which started in 2001, the performance test of the advanced spent fuel management process consisting of voloxidation, reduction of spent fuel and the lithium recovery process has been completed successfully on a laboratory scale. The world-premier spent fuel reduction hot test of a 5 kgHM/batch has been performed successfully by joint research with Russia and the valuable data on the actinides and FPs material balance and the characteristics of the metal product were obtained with experience to help design an engineering scale reduction system. The electrolytic reduction technology which integrates uranium oxide reduction in a molten LiCl-Li 2 O system and Li 2 O electrolysis is developed and a unique reaction system is also devised. Design data such as the treatment capacity, current density and mass transfer behavior obtained from the performance test of a 5 kgU/batch electrolytic reduction system pave the way for the third phase of the hot cell demonstration of the advanced spent fuel management technology
Transport and storage of spent nuclear fuel

International Nuclear Information System (INIS)

Lung, M.; Lenail, B.

1987-01-01

From a safety standpoint, spent fuel is clearly not ideal for permanent disposal and reprocessing is the best method of preparing wastes for long-term storage in a repository. Furthermore, the future may demonstrate that some fission products recovered in reprocessing have economic applications. Many countries have in fact reached the point at which the recycling of plutonium and uranium from spent fuel is economical in LWR's. Even in countries where this is not yet evident, (i.e., the United States), the French example shows that the day will come when spent fuel will be retrieved for reprocessing and recycle. It is highly questionable whether spent fuel will ever be considered and treated as waste in the same sense as fission products and processed as such, i.e., packaged in a waste form for permanent disposal. Even when recycled fuel material can no longer be reused in LWR's because of poor reactivity, it will be usable in FBR's. Based on the considerable experience gained by SGN and Cogema, this paper has provided practical discussion and illustrations of spent fuel transport and storage of a very important step in the nuclear fuel management process. The best of spent fuel storage depends on technical, economic and policy considerations. Each design has a role to play and we hope that the above discussion will help clarify certain issues
Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Process report

Energy Technology Data Exchange (ETDEWEB)

Gribi, Peter; Johnson, Lawrence; Suter, Daniel; Smith, Paul; Pastina, Barbara; Snellman, Margit

2008-01-15

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is placed in a perforated steel cylinder prior to emplacement; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the main processes potentially affecting the long-term safety of the system, covering radiation-related, thermal, hydraulic, mechanical, chemical (including microbiological) and radionuclide transport-related processes. The process descriptions deal sequentially with the main sub-systems: fuel/cavity in canister, cast iron insert and copper canister, buffer and other bentonite components, supercontainer
Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Process report

International Nuclear Information System (INIS)

Gribi, Peter; Johnson, Lawrence; Suter, Daniel; Smith, Paul; Pastina, Barbara; Snellman, Margit

2008-01-01

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is placed in a perforated steel cylinder prior to emplacement; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the main processes potentially affecting the long-term safety of the system, covering radiation-related, thermal, hydraulic, mechanical, chemical (including microbiological) and radionuclide transport-related processes. The process descriptions deal sequentially with the main sub-systems: fuel/cavity in canister, cast iron insert and copper canister, buffer and other bentonite components, supercontainer
Conditioning of spent nuclear fuel for permanent disposal

International Nuclear Information System (INIS)

Laidler, J.J.

1994-01-01

A compact, efficient method for conditioning spent nuclear fuel is under development This method, known as pyrochemical processing, or open-quotes pyroprocessing,close quotes provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (99.9%) separation of transuranics. The resultant waste forms from the pyroprocess are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and preclude the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory
Conditioning of spent nuclear fuel for permanent disposal

International Nuclear Information System (INIS)

Laidler, J.J.

1994-01-01

A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or pyroprocessing, provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the US Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (> 99.9%) separation of transuranics. The resultant waste forms from the pyroprocess are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and that avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory
Reactor-specific spent fuel discharge projections, 1987-2020

International Nuclear Information System (INIS)

Walling, R.C.; Heeb, C.M.; Purcell, W.L.

1988-03-01

The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs

Yugoslav spent nuclear fuel management program and international perspectives

International Nuclear Information System (INIS)

Pesic, M.; Subotic, K.; Sotic, O.; Plecas, I.; Ljubenov, V.; Peric, A.; Milosevic, M.

2002-01-01

Spent nuclear fuel stored in the Vinca Institute of Nuclear Sciences, Yugoslavia, consists of about 2.5 tons of metal uranium (initial enrichment 2%) and about 20 kg uranium dioxide (dispersed in aluminum matrix, initial fuel uranium enrichment 80%). This spent nuclear fuel is generated in operation of the RA heavy water research reactor during 1959-1984 period. Both types of fuel are of ex-USSR origin, have the same shape and dimensions and approximately the same initial mass of 235 nuclide. They are known as the TVR-S type of fuel elements. The total of 8030 spent fuel elements are stored at the RA research reactor premises, almost all in the spent fuel pool filled by ordinary water. The last used 480 high-enriched uranium spent fuel elements are kept in the drained RA reactor core since 1984. Fuel layer of both enrichments is covered with thin aluminium cladding. Due to non-suitable chemical parameters of water in the spent fuel storage pool, the corrosion processes penetrated aluminium cladding and aluminium walls od storage containers during storage period long from 20 to 40 years. Activity of fission products ( 137 Cs) is detected in water samples during water inspection in 1996 and experts of the lAEA Russia and USA were invited to help. By end of 2001, some remediation of the water transparency of the storage pool and inspections of water samples taken from the storage containers with the spent fuel elements were carried out by the Vinca Institute staff and with the help of experts from the Russia and the IAEA. Following new initiatives on international perspective on spent fuel management, a proposal was set by the IAEA, and was supported by the governments of the USA and the Russian Federation to ship the spent fuel elements of the RA research reactor to Mayak spent fuel processing plant in Russia. This paper describes current status of the reactor RA spent fuel elements, initiative for new Yugoslav spent fuel management program speculates on some of the
Spent-fuel storage: a private sector option

International Nuclear Information System (INIS)

Thomas, J.A.; Ross, S.R.

1983-01-01

The investigation was performed to delineate the legal and financial considerations for establishing private sector support for the planning and development of an independent spent-fuel storage facility (ISFSF). The preferred institutional structure was found to be one in which a not-for-profit corporation contracts with a limited partnership to handle the spent fuel. The limited partnership acquires the necessary land and constructs the ISFSF facility and then leases the facility to the not-for-profit corporation, which acquires spent-fuel rods from the utilities. The DOE must agree to purchase the spent-fuel rods at the expiration of term and warrant continued operation of the facility if policy changes at the federal level force the removal of the rods prior to completion of the contracted storage cycle. The DOE planning base estimate of spent-fuel storage requirements indicates a market potential adequate to support 10,000 MTU or more of spent-fuel storage prior to the time a government repository is available to accept spent fuel around the turn of the century. The estimated construction cost of a 5000-MTU water basin facility is $552 million. The total capital requirements to finance such a facility are estimated to be $695 million, based on an assumed capital structure of 70 percent debt and 30 percent equity. The estimated total levelized cost of storage, including operating costs, for the assumed 17-year life of the facility is $223 per kilogram of uranium. This is equivalent to a slightly less than one mill per kilowatt-hour increase in nuclear fuel costs at the nuclear power station that was the source of the spent fuel. In conclusion, within the context of the new Nuclear Waste Policy Act of 1982, the study points to both the need for and the advantages of private sector support for one or more ISFSFs and establishes a workable mechanism for the recovery of the costs of owning and operating such facilities. 3 figures, 4 tables
Development of advanced spent fuel management process. The fabrication and oxidation behavior of simulated metallized spent fuel

Energy Technology Data Exchange (ETDEWEB)

Ro, Seung Gy; Shin, Y.J.; You, G.S.; Joo, J.S.; Min, D.K.; Chun, Y.B.; Lee, E.P.; Seo, H.S.; Ahn, S.B

1999-03-01

The simulated metallized spent fuel ingots were fabricated and evaluated the oxidation rates and the activation energies under several temperature conditions to develop an advanced spent fuel management process. It was also checked the alloying characteristics of the some elements with metal uranium. (Author). 3 refs., 1 tab., 36 figs.
Emplacement of zero-valent metal for remediation of deep contaminant plumes

International Nuclear Information System (INIS)

Hubble, D.W.; Gillham, R.W.; Cherry, J.A.

1997-01-01

Some groundwater plumes containing chlorinated solvent contaminants are found to be so deep that current in situ remediation technologies cannot be economically applied. Also, source zones are often found to be too deep for removal or inaccessible due to surface features. Plumes emanating from these sources require containment or treatment. Containment technologies are available for shallow sites (< 15 m) and are being developed for greater depths. However, it is important to advance the science of reactive treatment - both for cut off of plumes and to contain and treat source zones. Zero-valent metal technology has been used for remediation of solvent plumes at sites in Canada, the UK and at several industrial and military sites in the USA. To date, all of the plumes treated with zero-valent metal (granular iron) have been at depths less than 15 m. This paper gives preliminary results of research into methods to emplace granular iron at depths in the range of 15 to 60 m. The study included review of available and emerging methods of installing barrier or reactive material and the selection, preliminary design and costing of several methods. The design of a treatment system for a 122 m wide PCE plume that, immediately down gradient from its source, extends from a depth of 24 to 37 m below the ground surface is used as a demonstration site. Both Permeable Reactive Wall and Funnel-and-Gate trademark systems were considered. The emplacement methods selected for preliminary design and costing were slurry wall, driven/vibrated beam, deep soil mixing and hydrofracturing injection. For each of these methods, the iron must be slurried for ease of pumping and placement using biodegradable polymer viscosifiers that leave the iron reactive
Long term wet spent nuclear fuel storage

International Nuclear Information System (INIS)

1987-04-01

The meeting showed that there is continuing confidence in the use of wet storage for spent nuclear fuel and that long-term wet storage of fuel clad in zirconium alloys can be readily achieved. The importance of maintaining good water chemistry has been identified. The long-term wet storage behaviour of sensitized stainless steel clad fuel involves, as yet, some uncertainties. However, great reliance will be placed on long-term wet storage of spent fuel into the future. The following topics were treated to some extent: Oxidation of the external surface of fuel clad, rod consolidation, radiation protection, optimum methods of treating spent fuel storage water, physical radiation effects, and the behaviour of spent fuel assemblies of long-term wet storage conditions. A number of papers on national experience are included
Effects of irregular basement structure on the geometry and emplacement of frontal thrusts and duplexes in the Quebec Appalachians: Interpretations from well and seismic reflection data

Science.gov (United States)

Konstantinovskaya, E.; Malo, M.; Badina, F.

2014-12-01

Irregular basement geometry may affect thrust propagation in foreland fold-thrust belts creating a perturbation in structural continuity of hydrocarbon fields. Here we investigate how the irregular pattern of normal faults, along with the presence of uplifts and transverse faults in the Grenvillian basement has influenced the geometry and emplacement of frontal thrusts and duplexes in the parautochthonous domain of the southern Quebec Appalachians during the middle-late Ordovician Taconian orogeny. Integration of data from surface geology, wells, and 2D and 3D seismic reflection surveys into a regional-scale structural model is used to reconstruct the 3D geometry and emplacement history of one- and two-horse duplexes in the Joly-Saint-Flavien gas storage area. The normal hinterland to foreland sequence of thrusting in this area is complicated by the differential emplacement of tectonic slices along strike of the orogenic front, starting in the SW and developing subsequently to the NE within each structural level. The shortening related to duplex emplacement decreases laterally over the distance of 20 km from - 49% in the SW (Joly area) to - 31% in the NE (Saint-Flavien area). Duplex emplacement resulted from the differential forward propagation, deflexion and vertical-axis rotation of the roof thrust (Logan's Line), which, in turn, has been induced by the presence of frontal uplift and transverse faults in the basement. The structural-lithological analysis of tectonic slices and restoration of their initial location allows us to consider the lower slice of the Joly duplex as a structural trap comparable to the fractured reservoir in lower Ordovician dolomites of the Saint-Flavien duplex.
Emplaced writing figuring as a manifestation of an ecocentric mindset in the narrative art of Petra Müller

Directory of Open Access Journals (Sweden)

Susan Meyer

2014-06-01

Full Text Available The human-earth connection is a sustained theme in Petra Müller’s oeuvre. The article focuses on this connection as reflected in her narrative art, specifically in accounts that have an autobiographical proclivity. The aim of this article is to outline the nature-centred disposition of Müller’s narrative art in a more definite sense. This is achieved by paying attention to the manner in which the author (the ‘I’ in accounts where the narrator can be identified as the author herself becomes part of the natural environment – whether on a sensory, an emotive, or an intellectual level – where she finds herself and the way she responds to it. At the core of the investigation are the ways in which this reactive engagement is manifested in Müller’s prose work by the implied author and the technique ofemplaced writing. Emplaced writing, a concept created by Linda Russo, was integrated by Susan Smith with Lawrence Buell’s concept of emplacement. This term refers to the technique allowing an active awareness of self and the place physically occupied by the author, as well as how that body fits into this place, to find expression. A broader perspective and greater appreciation of Müller’s work are drawn from the insight into how her close coexistence with the earth is reflected in her narrative art by means of the technique of emplaced writing which is explored in this article as it gives voice to a strong ecocentric disposition.
The Wall-Rock Record of Incremental Emplacement in the Little Cottonwood-Alta Magmatic and Hydrothermal System, Wasatch Mountains, Utah, U.S.A.

Science.gov (United States)

Stearns, M.; Callis, S.; Beno, C.; Bowman, J. R.; Bartley, J. M.

2017-12-01

Contact aureoles record the cumulative effects on wall rocks of magma emplacement. Like the plutons they surround, contact aureoles have long been regarded to form geologically instantaneously. Protracted incremental emplacement of plutons must be reconciled with the wall-rock record of heat and mass transfer. Fundamental questions include how heat and material move from intrusions into their aureoles and how long that process takes. The Little Cottonwood stock is surrounded by a 2 km-wide contact aureole that contains prograde AFM mineral assemblages in the pelitic layers of the Proterozoic Big Cottonwood Formation. The Alta stock is surrounded by a well characterized 1 km-wide contact aureole containing both prograde AFM and CMS mineral assemblages in Ophir Shale and Mississippian dolostones, respectively. Understanding the petrogenesis of these aureoles requires the timing of magmatism and wall-rock metamorphism to be independently determined. Preliminary petrochronology (U/Th-Pb dates and trace element concentrations collected by LASS-ICP-MS) from the inner aureoles of both intrusions establishes a protracted history of monazite (re)crystallization from 35-25 Ma in the Little Cottonwood aureole and 35 Ma in the Alta aureole. Little Cottonwood aureole monazites are characterized by a positive age correlation with heavy rare earth elements (HREE) and a negative correlation with Eu/Eu*. Alta aureole monazites have a similar range of the HREE concentrations and Eu/Eu* variation. Zircon growth interpreted to record emplacement-level magmatic crystallization of the western Little Cottonwood stock ranges from 33-28 Ma near the contact. Multi-grain U-Pb zircon TIMS dates from the Alta stock range from 35-33 Ma and are interpreted to suggest the full range of emplacement-level magmatism in the Alta stock. Additionally, in situ U-Pb titanite dates from the Alta stock record intermittent high temperature hydrothermal activity in the stock margin from 35-24 Ma. These new
Transportation of spent MTR fuels

Energy Technology Data Exchange (ETDEWEB)

Raisonnier, D.

1997-08-01

This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs.
Transportation of spent MTR fuels

International Nuclear Information System (INIS)

Raisonnier, D.

1997-01-01

This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs
The Jeffers Brook diorite-granodiorite pluton: style of emplacement and role of volatiles at various crustal levels in Avalonian appinites, Canadian Appalachians

Science.gov (United States)

Pe-Piper, Georgia; Piper, David J. W.

2018-04-01

Small appinite plutons ca. 610 Ma outcrop in the peri-Gondwanan Avalon terrane of northern Nova Scotia, with different structural levels exposed. Field mapping shows that the Jeffers Brook pluton is a laccolith emplaced along an upper crustal thrust zone, likely in a dilational jog in a regional dextral strike-slip system. The oldest rocks are probably mafic sills, which heated the area facilitating emplacement of intermediate magmas. Cross-cutting relationships show that both mafic and intermediate magmas were supplied throughout the history of pluton emplacement. The modal composition, mineral chemistry, and bulk chemistry of gabbro, diorite, tonalite, granodiorite, and granite have been studied in the main plutonic phases, dykes, and sills, and mafic microgranular enclaves. As with the type appinites in the Scottish Caledonides, the pluton shows evidence of high water content: the dominance of hornblende, locally within pegmatitic texture; vesicles and irregular felsic patches in enclaves; and late aplite dykes. Analyzed mafic microgranular enclaves are geochemically similar to larger diorite bodies in the pluton. Tonalite-granodiorite is distinct from the diorite in trace-element geochemistry and radiogenic isotopes. Elsewhere to the east, similar rocks of the same age form vertically sheeted complexes in major shear zones; hornblende chemistry shows that they were emplaced at a deeper upper crustal level. This implies that little of the observed geochemical variability in the Jeffers Brook pluton was developed within the pluton. The general requirements to form appinites are proposed to be small magma volumes of subduction-related magmas that reach the upper crust because of continual heating by mafic magmas moving through strike-slip fault pathways and trapping of aqueous fluids rather than venting through volcanic activity.
Operation of spent fuel storage facilities

International Nuclear Information System (INIS)

1994-01-01

This Safety Guide was prepared as part of the IAEA's programme on safety of spent fuel storage. This is for interim spent fuel storage facilities that are not integral part of an operating nuclear power plant. Following the introduction, Section 2 describes key activities in the operation of spent fuel storage facilities. Section 3 lists the basic safety considerations for storage facility operation, the fundamental safety objectives being subcriticality, heat removal and radiation protection. Recommendations for organizing the management of a facility are contained in Section 4. Section 5 deals with aspects of training and qualification; Section 6 describes the phases of the commissioning of a spent fuel storage facility. Section 7 describes operational limits and conditions, while Section 8 deals with operating procedures and instructions. Section 9 deals with maintenance, testing, examination and inspection. Section 10 presents recommendations for radiation and environmental protection. Recommendations for the quality assurance (QA) system are presented in Section 11. Section 12 describes the aspects of safeguards and physical protection to be taken into account during operations; Section 13 gives guidance for decommissioning. 15 refs, 5 tabs
Preliminary Calculation on a Spent Fuel Pool Accident using GOTHIC

Energy Technology Data Exchange (ETDEWEB)

Park, Jaehwan; Choi, Yu Jung; Hong, Tae Hyub; Kim, Hyeong-Taek [KHNP-CRI, Daejeon (Korea, Republic of)

2015-10-15

The probability of an accident happening at the spent fuel pool was believed to be quite low until the 2011 Fukushima accident occurred. Notably, large amount of spent fuel are normally stored in the spent fuel pool for a long time compared to the amount of fuel in the reactor core and the total heat released from the spent fuel is high enough to boil the water of the spent fuel pool when the cooling system does not operate. In addition, the enrichment and the burnup of the fuel have both increased in the past decade and heat generation from the spent fuel thereby has also increased. The failure of the cooling system at the spent fuel pool (hereafter, a loss-of-cooling accident) is one of the principal hypothetical causes of an accident that could occur at the spent fuel pool. In this paper, the preliminary calculation of a loss-of-cooling accident was performed. In this paper, the preliminary calculation of a loss-of cooling accident was performed with GOTHIC. The calculation results show boiling away of water in the spent fuel pool due to the loss-of-cooling accident and similar thermal performance of the spent fuel pool with previous research results.
Optimization of time and location dependent spent nuclear fuel storage capacity

International Nuclear Information System (INIS)

Macek, V.

1977-01-01

A linear spent fuel storage model is developed to identify cost-effective spent nuclear fuel storage strategies. The purpose of this model is to provide guidelines for the implementation of the optimal time-dependent spent fuel storage capacity expansion in view of the current economic and regulatory environment which has resulted in phase-out of the closed nuclear fuel cycle. Management alternatives of the spent fuel storage backlog, which is created by mismatch between spent fuel generation rate and spent fuel disposition capability, are represented by aggregate decision variables which describe the time dependent on-reactor-site and off-site spent fuel storage capacity additions, and the amount of spent fuel transferred to off-site storage facilities. Principal constraints of the model assure determination of cost optimal spent fuel storage expansion strategies, while spent fuel storage requirements are met at all times. A detailed physical and economic analysis of the essential components of the spent fuel storage problem, which precedes the model development, assures its realism. The effects of technological limitations on the on-site spent fuel storage expansion and timing of reinitiation of the spent fuel reprocessing on optimal spent fuel storage capacity expansion are investigated. The principal results of the study indicate that (a) expansion of storage capacity beyond that of currently planned facilities is necessary, and (b) economics of the post-reactor fuel cycle is extremely sensitive to the timing of reinitiation of spent fuel reprocessing. Postponement of reprocessing beyond mid-1982 may result in net negative economic liability of the back end of the nuclear fuel cycle
Safe transport of spent fuels after long-term storage

International Nuclear Information System (INIS)

Aritomi, M.; Takeda, T.; Ozaki, S.

2004-01-01

Considering the scarcity of energy resources in Japan, a nuclear energy policy pertaining to the spent fuel storage has been adopted. The nuclear energy policy sets the rules that spent fuels generated from LWRs shall be reprocessed and that plutonium and unburnt uranium shall be recovered and reused. For this purpose, a reprocessing plant, which has a reprocessing capability of 800 ton/yr, is under construction at Rokkasho Village. However, it is anticipated that the start of its operation will be delayed. In addition, the amount of spent fuels generated from nuclear power plants exceeds its reprocessing capability. Therefore, the establishment of storage technology for spent fuels becomes an urgent problem in Japan in order to continue smoothly the LWR operations. In this paper, the background of nuclear power generation in Japan is introduced at first. Next, the policy of spent fuel storage in Japan and circumstances surrounding the spent fuels in Japan are mentioned. Furthermore, the major subjects for discussions to settle and improve 'Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facility' in Atomic Energy Society of Japan are discussed, such as the integrity of fuel cladding, basket, shielding material and metal gasket for the long term storage for achieving safe transport of spent fuels after the storage. Finally, solutions to the unsolved subject in establishing the spent fuel interim storage technologies ase introduced accordingly
Laser surveillance system for spent fuel

International Nuclear Information System (INIS)

Fiarman, S.; Zucker, M.S.; Bieber, A.M. Jr.

1980-01-01

A laser surveillance system installed at spent fuel storage pools will provide the safeguard inspector with specific knowledge of spent fuel movement that cannot be obtained with current surveillance systems. The laser system will allow for the division of the pool's spent fuel inventory into two populations - those assemblies which have been moved and those which haven't - which is essential for maximizing the efficiency and effectiveness of the inspection effort. We have designed, constructed, and tested a laser system and have used it with a simulated BWR assembly. The reflected signal from the zircaloy rods depends on the position of the assembly, but in all cases is easily discernable from the reference scan of background with no assembly
The Canadian research reactor spent fuel situation

International Nuclear Information System (INIS)

Ernst, P.C.

1996-01-01

This paper summarizes the present research reactor spent fuel situation in Canada. The research reactors currently operating are listed along with the types of fuel that they utilize. Other shut down research reactors contributing to the storage volume are included for completeness. The spent fuel storage facilities associated with these reactors and the methods used to determine criticality safety are described. Finally the current inventory of spent fuel and where it is stored is presented along with concerns for future storage. (author). 3 figs
Investigation of the condition of spent-fuel pool components

International Nuclear Information System (INIS)

Kustas, F.M.; Bates, S.O.; Opitz, B.E.; Johnson, A.B. Jr.; Perez, J.M. Jr.; Farnsworth, R.K.

1981-09-01

It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts
Investigation of the condition of spent-fuel pool components

Energy Technology Data Exchange (ETDEWEB)

Kustas, F.M.; Bates, S.O.; Opitz, B.E.; Johnson, A.B. Jr.; Perez, J.M. Jr.; Farnsworth, R.K.

1981-09-01

It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts.
Manufacturing and Construction of Spent Fuel Storage Rack for Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Sangjin; Jung, Kwangsub; Oh, Jinho; Lee, Jongmin [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The spent fuel storage rack consists of spent fuel storage racks and support frame. The spent fuel storage racks are installed in the support frame. A spent fuel storage rack consists of frame weldment and storage cell pipe assembly. Storage cell pipe assembly is mounted on the base plate of the frame weldment. The spent fuel storage rack is designed to withstand seismic load and other loads during earthquake. The structural integrity of the spent fuel storage rack is evaluated in accordance with ASME Section III, Subsection NF. Computer Code used for this analysis is ANSYS version 14.0.0. Dead load and seismic load is considered in load condition and hydrodynamic mass is included in the analysis. Design, manufacturing, and construction of the spent fuel storage rack are introduced. The spent fuel storage rack is for storage of spent fuel assemblies. The spent fuel storage rack should be designed, manufactured, and installed with consideration of predicted number of spent fuel assemblies, structural integrity, resistivity to corrosion and radiation, cleaning, and workability.

Paleomagnetism in the Determination of the Emplacement Temperature of Cerro Colorado Tuff Cone, El Pinacate Volcanic Field, Sonora, Mexico.

Science.gov (United States)

Rodriguez Trejo, A.; Alva-Valdivia, L. M.; Vidal Solano, J. R.; Garcia Amador, B.; Gonzalez-Rangel, J. A.

2014-12-01

Cerro Colorado Maar is located at the World Heritage Site, biosphere reserve El Pinacate and Gran Desierto del Altar, at the NNW region of Sonora, Mexico (in El Pinacate Volcanic Field). It is a tuff cone, about 1 km diameter, result of several phreatomagmatic episodes during the late Quaternary. We report paleomagnetic and rock magnetic properties from fusiform volcanic bombs obtained from the borders of Cerro Colorado. This study is based in the thermoremanent magnetization TRM normally acquired by volcanic rocks, which can be used to estimate the emplacement temperature range. We performed the experiments on 20 lithic fragments (10 cm to 20 cm approximately), taking 6-8 paleomagnetic cores from each. Rock magnetic experiments (magnetic susceptibility vs. temperature (k-T), hysteresis curves and FORC analysis, shows that the main magnetic mineral carriers of magnetization are titanomagnetite and titanohematite in different levels of intergrowth. The k-T curves suggest in many cases, only one magnetic phase, but also in other cases a second magnetic phase. Thermal demagnetization was used to demagnetize the specimens in detailed short steps and make a well-defined emplacement temperature determination ranges. We found that temperature emplacement determination range for these two magnetic phases is between 350-450 °C, and 550-580 °C, respectively. These results are consistent with those expected in an eruption of Surtsey type, showing a distinct volcanic activity compared to the other craters from El Pinacate volcanic field.
Tracking lava flow emplacement on the east rift zone of Kīlauea, Hawai‘i, with synthetic aperture radar coherence

Science.gov (United States)

Dietterich, Hannah R.; Poland, Michael P.; Schmidt, David A.; Cashman, Katharine V.; Sherrod, David R.; Espinosa, Arkin Tapia

2012-05-01

Lava flow mapping is both an essential component of volcano monitoring and a valuable tool for investigating lava flow behavior. Although maps are traditionally created through field surveys, remote sensing allows an extraordinary view of active lava flows while avoiding the difficulties of mapping on location. Synthetic aperture radar (SAR) imagery, in particular, can detect changes in a flow field by comparing two images collected at different times with SAR coherence. New lava flows radically alter the scattering properties of the surface, making the radar signal decorrelated in SAR coherence images. We describe a new technique, SAR Coherence Mapping (SCM), to map lava flows automatically from coherence images independent of look angle or satellite path. We use this approach to map lava flow emplacement during the Pu`u `Ō`ō-Kupaianaha eruption at Kīlauea, Hawai`i. The resulting flow maps correspond well with field mapping and better resolve the internal structure of surface flows, as well as the locations of active flow paths. However, the SCM technique is only moderately successful at mapping flows that enter vegetation, which is also often decorrelated between successive SAR images. Along with measurements of planform morphology, we are able to show that the length of time a flow stays decorrelated after initial emplacement is linearly related to the flow thickness. Finally, we use interferograms obtained after flow surfaces become correlated to show that persistent decorrelation is caused by post-emplacement flow subsidence.
Tracking lava flow emplacement on the east rift zone of Kilauea, Hawai’i with synthetic aperture radar (SAR) coherence

Science.gov (United States)

Dietterich, Hannah R.; Poland, Michael P.; Schmidt, David; Cashman, Katharine V.; Sherrod, David R.; Espinosa, Arkin Tapia

2012-01-01

Lava flow mapping is both an essential component of volcano monitoring and a valuable tool for investigating lava flow behavior. Although maps are traditionally created through field surveys, remote sensing allows an extraordinary view of active lava flows while avoiding the difficulties of mapping on location. Synthetic aperture radar (SAR) imagery, in particular, can detect changes in a flow field by comparing two images collected at different times with SAR coherence. New lava flows radically alter the scattering properties of the surface, making the radar signal decorrelated in SAR coherence images. We describe a new technique, SAR Coherence Mapping (SCM), to map lava flows automatically from coherence images independent of look angle or satellite path. We use this approach to map lava flow emplacement during the Pu‘u ‘Ō‘ō-Kupaianaha eruption at Kīlauea, Hawai‘i. The resulting flow maps correspond well with field mapping and better resolve the internal structure of surface flows, as well as the locations of active flow paths. However, the SCM technique is only moderately successful at mapping flows that enter vegetation, which is also often decorrelated between successive SAR images. Along with measurements of planform morphology, we are able to show that the length of time a flow stays decorrelated after initial emplacement is linearly related to the flow thickness. Finally, we use interferograms obtained after flow surfaces become correlated to show that persistent decorrelation is caused by post-emplacement flow subsidence.
Choosing a spent fuel interim storage system

International Nuclear Information System (INIS)

Roland, V.; Hunter, I.

2001-01-01

The Transnucleaire Group has developed different modular solutions to address spent fuel interim storage needs of NPP. These solutions, that are present in Europe, USA and Asia are metal casks (dual purpose or storage only) of the TN 24 family and the NUHOMS canister based system. It is not always simple for an operator to sort out relevant choice criteria. After explaining the basic designs involved on the examples of the TN 120 WWER dual purpose cask and the NUHOMS 56 WWER for WWER 440 spent fuel, we shall discuss the criteria that govern the choice of a given spent fuel interim storage system from the stand point of the operator. In conclusion, choosing and implementing an interim storage system is a complex process, whose implications can be far reaching for the long-term success of a spent fuel management policy. (author)
Development and engineering plan for graphite spent fuels conditioning program

International Nuclear Information System (INIS)

Bendixsen, C.L.; Fillmore, D.L.; Kirkham, R.J.; Lord, D.L.; Phillips, M.B.; Pinto, A.P.; Staiger, M.D.

1993-09-01

Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated
The EBR-II spent fuel treatment program

International Nuclear Information System (INIS)

Lineberry, M.J.; McFarlane, H.F.

1995-01-01

Argonne National Laboratory has refurbished and equipped an existing hot cell facility for demonstrating a high-temperature electrometallurgical process for treating spent nuclear fuel from the Experimental Breeder Reactor-11. Two waste forms will be produced and qualified for geologic disposal of the fission and activation products. Relatively pure uranium will be separated for storage. Following additional development, transuranium elements will be blended into one of the high-level waste streams. The spent fuel treatment program will help assess the viability of electrometallurgical technology as a spent fuel management option
Advances in HTGR spent fuel treatment technology

International Nuclear Information System (INIS)

Holder, N.D.; Lessig, W.S.

1984-08-01

GA Technologies, Inc. has been investigating the burning of spent reactor graphite under Department of Energy sponsorship since 1969. Several deep fluidized bed burners have been used at the GA pilot plant to develop graphite burning techniques for both spent fuel recovery and volume reduction for waste disposal. Since 1982 this technology has been extended to include more efficient circulating bed burners. This paper includes updates on high-temperature gas-cooled reactor fuel cycle options and current results of spent fuel treatment testing for fluidized and advanced circulating bed burners
The management status of the spent fuel in HANARO(1995-2009)

International Nuclear Information System (INIS)

Choi, Ho Young; Lim, Kyeng Hwan; Kim, Hyung Wook; Lee, Choong Sung; Ahn, Guk Hoon

2009-11-01

In HANARO, the spent fuels are stored in the spent fuel storage pool of the reactor hall. The capacity of the spent fuel storage pool was designed to store 600 bundles for 36 rods fuel, 432 bundles for 18 rods fuel, 315 rods for TRIGA reactor fuel and the fuels loaded in the reactor core. The spent fuel storage pool can store spent fuels discharged from the reactor core for 20 years normal operation. As for July 2009, the spent fuel 337 bundles are stored in the spent fuel storage pool. There are 217 bundles of 36 rods fuel and 120 bundles of 18 rods fuel. In this report, the information of the spent fuel about the loading date in the reactor core, discharged date, burnup, invisible inspection results and loading position in the spent fuel storage pool are described
Spent nuclear fuel in Bulgaria

International Nuclear Information System (INIS)

Peev, P.; Kalimanov, N.

1999-01-01

The development of the nuclear energy sector in Bulgaria is characterized by two major stages. The first stage consisted of providing a scientific basis for the programme for development of the nuclear energy sector in the country and was completed with the construction of an experimental water-water reactor. At present, spent nuclear fuel from this reactor is placed in a water filled storage facility and will be transported back to Russia. The second stage consisted of the construction of the 6 NPP units at the Kozloduy site. The spent nuclear fuel from the six units is stored in at reactor pools and in an additional on-site storage facility which is nearly full. In order to engage the government of the country with the on-site storage problems, the new management of the National Electric Company elaborated a policy on nuclear fuel cycle and radioactive waste management. The underlying policy is de facto the selection of the 'deferred decision' option for its spent fuel management. (author)
Spent fuel management in Japan

International Nuclear Information System (INIS)

Shirahashi, K.; Maeda, M.; Nakai, T.

1996-01-01

Japan has scarce energy resources and depends on foreign resources for 84% of its energy needs. Therefore, Japan has made efforts to utilize nuclear power as a key energy source since mid-1950's. Today, the nuclear energy produced from 49 nuclear power plants is responsible for about 31% of Japan's total electricity supply. The cumulative amount of spent fuel generated as of March 1995 was about 11,600 Mg U. Japan's policy of spent fuel management is to reprocess spent nuclear fuel and recycle recovered plutonium and uranium as nuclear fuel. The Tokai reprocessing plant continues stable operation keeping the annual treatment capacity or around 90 Mg U. A commercial reprocessing plant is under construction at Rokkasho, northern part of Japan. Although FBR is the principal reactor to use plutonium, LWR will be a major power source for some time and recycling of the fuel in LWRs will be prompted. (author). 3 figs
Spent nuclear fuel storage vessel

International Nuclear Information System (INIS)

Watanabe, Yoshio; Kashiwagi, Eisuke; Sekikawa, Tsutomu.

1997-01-01

Containing tubes for containing spent nuclear fuels are arranged vertically in a chamber. Heat releasing fins are disposed horizontal to the outer circumference of the containing tubes for rectifying cooling air and promoting cooling of the containing tubes. Louvers and evaporation sides of heat pipes are disposed at a predetermined distance in the chamber. Cooling air flows from an air introduction port to the inside of the chamber and takes heat from the containing tubes incorporated with heat generating spent nuclear fuels, rising its temperature and flows off to an air exhaustion exit. The direction for the rectification plate of the louver is downward from a horizontal position while facing to the air exhaustion port. Since the evaporation sides of the heat pipes are disposed in the inside of the chamber and the condensation side of the heat pipes is disposed to the outside of the chamber, the thermal energy can be recovered from the containing tubes incorporated with spent nuclear fuels and utilized. (I.N.)
Features and safety aspects of spent fuel storage facility, Tarapur

International Nuclear Information System (INIS)

Pradhan, Sanjay; Dubey, K.; Qureshi, F.T.; Lokeswar, S.P.

2017-01-01

Spent Fuel Storage Facility (SFSF), Tarapur is designed to store spent fuel arising from PHWRs in different parts of the country. Spent fuel is transported in AERB qualified/authorized shipping cask by NPCIL to SFSF by road or rail route. The spent fuel storage facility at Tarapur was hot commissioned after regulatory clearances
A present status for dry storage of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Bang, K. S.; Lee, J. C.; Park, H. Y.; Seo, K. S

2003-04-01

National policy for management of a spent nuclear fuel does not establish in Korea yet. A storage capacity of a storage pool that is to store the spent nuclear fuel will be exceeded an amount of accumulation from the first Woljin nuclear power plant in 2007. Therefore it is necessary that dry storage facility is secured to store safely the spent nuclear fuel on site of the nuclear power plant until national policy for a back-end spent nuclear fuel cycle is established. In order to store safely spent nuclear fuel, it is important that the present status and technology on dry storage of spent nuclear fuel is looked over. Therefore, the present status on dry storage of spent nuclear fuel was analyzed so as to develop dry storage system and choose a proper dry storage method domestic.
Investigation of novel spent fuel verification system for safeguard application

Energy Technology Data Exchange (ETDEWEB)

Lee, Haneol; Yim, Man-Sung [KAIST, Daejeon (Korea, Republic of)

2016-10-15

Radioactive waste, especially spent fuel, is generated from the operation of nuclear power plants. The final stage of radioactive waste management is disposal which isolates radioactive waste from the accessible environment and allows it to decay. The safety, security, and safeguard of a spent fuel repository have to be evaluated before its operation. Many researchers have evaluated the safety of a repository. These researchers calculated dose to public after the repository is closed depending on their scenario. Because most spent fuel repositories are non-retrievable, research on security or safeguards of spent fuel repositories have to be performed. Design based security or safeguard have to be developed for future repository designs. This study summarizes the requirements of future spent fuel repositories especially safeguards, and suggests a novel system which meets the safeguard requirements. Applying safeguards to a spent fuel repository is becoming increasingly important. The future requirements for a spent fuel repository are suggested by several expert groups, such as ASTOR in IAEA. The requirements emphasizes surveillance and verification. The surveillance and verification of spent fuel is currently accomplished by using the Cerenkov radiation detector while spent fuel is being stored in a fuel pool. This research investigated an advanced spent fuel verification system using a system which converts spent fuel radiation into electricity. The system generates electricity while it is conveyed from a transportation cask to a disposal cask. The electricity conversion system was verified in a lab scale experiment using an 8.51GBq Cs-137 gamma source.
Investigation of novel spent fuel verification system for safeguard application

International Nuclear Information System (INIS)

Lee, Haneol; Yim, Man-Sung

2016-01-01

Radioactive waste, especially spent fuel, is generated from the operation of nuclear power plants. The final stage of radioactive waste management is disposal which isolates radioactive waste from the accessible environment and allows it to decay. The safety, security, and safeguard of a spent fuel repository have to be evaluated before its operation. Many researchers have evaluated the safety of a repository. These researchers calculated dose to public after the repository is closed depending on their scenario. Because most spent fuel repositories are non-retrievable, research on security or safeguards of spent fuel repositories have to be performed. Design based security or safeguard have to be developed for future repository designs. This study summarizes the requirements of future spent fuel repositories especially safeguards, and suggests a novel system which meets the safeguard requirements. Applying safeguards to a spent fuel repository is becoming increasingly important. The future requirements for a spent fuel repository are suggested by several expert groups, such as ASTOR in IAEA. The requirements emphasizes surveillance and verification. The surveillance and verification of spent fuel is currently accomplished by using the Cerenkov radiation detector while spent fuel is being stored in a fuel pool. This research investigated an advanced spent fuel verification system using a system which converts spent fuel radiation into electricity. The system generates electricity while it is conveyed from a transportation cask to a disposal cask. The electricity conversion system was verified in a lab scale experiment using an 8.51GBq Cs-137 gamma source
Engineering studies: high-level radioactive waste repositories task 3 - review of underground handling and emplacement. 1. Executive summary

Energy Technology Data Exchange (ETDEWEB)

1981-02-01

The report reviews proposals for transport, handling and emplacement of high-level radioactive waste in an underground repository with particular reference to: waste block size and configuration; self-shielded or partially-shielded block; stages of disposal; transport by road/rail to repository site; handling techniques within repository; emplacement in vertical holes or horizontal tunnels; repository access by adit, incline or shaft; conventional and radiological safety; costs; and major areas of uncertainty requiring research or development. In carrying out this programme due attention was given to work already carried out both in the U.K. and overseas and where appropriate comparisons with this work have been made to substantiate and explain the observations made in this report. The examination and use of this previous work however has been confined to those proposals which were considered capable of meeting the basic design criterion for a U.K. based repository, that the maximum temperature achieved by the host rock should not exceed 100/sup 0/C.
Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, vertical emplacement mode: Volume 1

International Nuclear Information System (INIS)

1987-12-01

This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packing, and other surface facility operations are described. Operations in the shafts underground are described, including waste hoisting, transfer, and vertical emplacement. This report specifically addresses the vertical emplacement mode, the reference design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs
Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, vertical emplacement mode: Volume 1

Energy Technology Data Exchange (ETDEWEB)

1987-12-01

This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packing, and other surface facility operations are described. Operations in the shafts underground are described, including waste hoisting, transfer, and vertical emplacement. This report specifically addresses the vertical emplacement mode, the reference design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs.
Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, horizontal emplacment mode: Volume 1

International Nuclear Information System (INIS)

1987-12-01

This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packaging, and other surface facility operations are described. Operations in the shafts and underground are described, including waste hoisting, transfer, and horizontal emplacement. This report specifically addresses the horizontal emplacement mode, the passive alternate design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs
Evolution of spent fuel dry storage

Energy Technology Data Exchange (ETDEWEB)

Standring, Paul Nicholas [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Fuel Cycle and Waste Technology; Takats, Ferenc [TS ENERCON KFT, Budapest (Hungary)

2016-11-15

Around 10,000 tHM of spent fuel is discharged per year from the nuclear power plants in operation. Whilst the bulk of spent fuel is still held in at reactor pools, 24 countries have developed storage facilities; either on the reactor site or away from the reactor site. Of the 146 operational AFR storage facilities about 80 % employ dry storage; the majority being deployed over the last 20 years. This reflects both the development of dry storage technology as well as changes in politics and trading relationships that have affected spent fuel management policies. The paper describes the various approaches to the back-end of the nuclear fuel cycle for power reactor fuels and provides data on deployed storage technologies.

Spent fuel storage requirements 1989--2020

International Nuclear Information System (INIS)

1989-10-01

Historical inventories of spent fuel are combined with Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the US to provide estimates of spent fuel storage requirements over the next 32 years, through the year 2020. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Historical data through December 1988 are derived from the 1989 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 14 refs., 3 figs., 28 tabs
Thermal-hydraulic analysis of spent fuel storage systems

International Nuclear Information System (INIS)

Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

1987-01-01

This paper describes the COBRA-SFS (Spent Fuel Storage) computer code, which is designed to predict flow and temperature distributions in spent nuclear fuel storage and transportation systems. The decay heat generated by spent fuel in a dry storage cask is removed through a combination of conduction, natural convection, and thermal radiation. One major advantage of COBRA-SFS is that fluid recirculation within the cask is computed directly by solving the mass and momentum conservation equations. In addition, thermal radiation heat transfer is modeled using detailed radiation exchange factors based on quarter-rod segments. The equations governing mass, momentum, and energy conservation for incompressible flows are presented, and the semi-implicit solution method is described. COBRA-SFS predictions are compared to temperature data from a spent fuel storage cask test and the effect of different fill media on the cladding temperature distribution is discussed. The effect of spent fuel consolidation on cask thermal performance is also investigated. 16 refs., 6 figs., 2 tabs
Cask operation and maintenance for spent fuel storage

Energy Technology Data Exchange (ETDEWEB)

Lee, J.S. [International Atomic Energy Agency, Vienna (Austria)

2004-07-01

Interim storage is an essential platform for any option to be chosen later as an endpoint for spent fuel management. In view of such a circumstance, the most imminent service required for the spent fuel management worldwide is to provide adequate storage for the future spent fuel inventory arising either from the continued operation of nuclear power plants or from the removal of spent fuel in preparation for plant decommissioning. While the bulk of the global inventory of spent fuel are still stored in AR pools, dry storage has become a prominent alternative especially for newly built AFR facilities, with more than 17,000 t HM already stored in dry storage facilities worldwide. Storage in cask under inert conditions has become the preferred option, given the advantages including passive cooling features and modular mode of capacity increase. In terms of economics, dry storage is particularly propitious for long-term storage in that operational costs are minimized by the passive cooling features. The trend toward dry storage, especially in cask type, is likely to continue with an implication that and the supply will closely follow the increasing demand for storage by incremental additions of casks to the effect of minimizing cost penalty of the idle capacities typical of pool facilities. A variety of storage systems have been developed to meet specific requirements of different reactor fuels and a large number of designs based on these generic technologies are now available for the spent fuel containers (horizontal, vertical etc) and storage facilities. Multi-purpose technologies (i.e. a single technology for storage, transportation and disposal) have also been studied. Recent concern on security measures for protection of spent fuel has prompted a consideration on the possibility of placing storage facility underground. The future evolution of requirements and technologies will bring important impacts on cask operation and maintenance for spent fuel storage.
Cask operation and maintenance for spent fuel storage

International Nuclear Information System (INIS)

Lee, J.S.

2004-01-01

Interim storage is an essential platform for any option to be chosen later as an endpoint for spent fuel management. In view of such a circumstance, the most imminent service required for the spent fuel management worldwide is to provide adequate storage for the future spent fuel inventory arising either from the continued operation of nuclear power plants or from the removal of spent fuel in preparation for plant decommissioning. While the bulk of the global inventory of spent fuel are still stored in AR pools, dry storage has become a prominent alternative especially for newly built AFR facilities, with more than 17,000 t HM already stored in dry storage facilities worldwide. Storage in cask under inert conditions has become the preferred option, given the advantages including passive cooling features and modular mode of capacity increase. In terms of economics, dry storage is particularly propitious for long-term storage in that operational costs are minimized by the passive cooling features. The trend toward dry storage, especially in cask type, is likely to continue with an implication that and the supply will closely follow the increasing demand for storage by incremental additions of casks to the effect of minimizing cost penalty of the idle capacities typical of pool facilities. A variety of storage systems have been developed to meet specific requirements of different reactor fuels and a large number of designs based on these generic technologies are now available for the spent fuel containers (horizontal, vertical etc) and storage facilities. Multi-purpose technologies (i.e. a single technology for storage, transportation and disposal) have also been studied. Recent concern on security measures for protection of spent fuel has prompted a consideration on the possibility of placing storage facility underground. The future evolution of requirements and technologies will bring important impacts on cask operation and maintenance for spent fuel storage
Management of spent sealed sources in Indonesia

International Nuclear Information System (INIS)

Wisnubroto, D.S.

2002-01-01

This paper describes the effort of the Center for Development of Radioactive Waste Management (CDRWM) to develop and implement activities in maintaining and improving the safety of spent sealed radiation sources and the security of radioactive materials over their life cycle. There is a wide variety of uses of radiation sources and radioactive materials in Indonesia, while the CDRWM plan to cover all spent radiation sources. Primary consideration is given to sealed radiation sources with relatively high levels of radioactivity which might necessitate interventional measures should control over them be lost. The policy of the Government of Indonesia for spent radiation sources is, whenever possible, spent sealed sources should be returned to the supplier. CDRWM has a general principle that sealed sources should not be removed from their holders, or the holders physically modified (except for Ra-226 needles, smoke detector and lighting preventer). (author)
International experience in conditioning spent fuel elements

International Nuclear Information System (INIS)

Ashton, P.

1991-04-01

The purpose of this report is to compile and present in a clear form international experience (USA, Canada, Sweden, FRG, UK, Japan, Switzerland) gained to date in conditioning spent fuel elements. The term conditioning is here taken to mean the handling and packaging of spent fuel elements for short- or long-term storage or final disposal. Plants of a varying nature fall within this scope, both in terms of the type of fuel element treated and the plant purpose eg. experimental or production plant. Emphasis is given to plants which bear some similarity to the concept developed in Germany for direct disposal of spent fuel elements. Worldwide, however, relatively few conditioning plants are in existence or have been conceived. Hence additional plants have been included where aspects of the experience gained are also of relevance eg. plants developed for the consolidation of spent fuel elements. (orig./HP) [de
TRIGA Mark II Ljubljana - spent fuel transportation

International Nuclear Information System (INIS)

Ravnik, M.; Dimic, V.

2008-01-01

The most important activity in 1999 was shipment of the spent fuel elements back to the United States for final disposal. This activity started already in 1998 with some governmental support. In July 1999 all spent fuel elements (219 pieces) from the TRIGA research reactor in Ljubljana were shipped back to the United Stated by the ship from the port Koper in Slovenia. At the same time shipment of the spent fuel from the research reactor in Pitesti, Romania, and the research reactor in Rome, Italy, was conducted. During the loading the radiation exposure to the workers was rather low. The loading and shipment of the spent nuclear fuel went very smoothly and according the accepted time table. During the last two years the TRIGA research reactor in Ljubljana has been in operation about 1100 hours per year and without any undesired shut-down. (authors)
Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes

Energy Technology Data Exchange (ETDEWEB)

Grazevicius, Audrius; Kaliatka, Algirdas [Lithuanian Energy Institute, Kaunas (Lithuania). Lab. of Nuclear Installation Safety

2017-07-15

The main functions of spent fuel pools are to remove the residual heat from spent fuel assemblies and to perform the function of biological shielding. In the case of loss of heat removal from spent fuel pool, the fuel rods and pool water temperatures would increase continuously. After the saturated temperature is reached, due to evaporation of water the pool water level would drop, eventually causing the uncover of spent fuel assemblies, fuel overheating and fuel rods failure. This paper presents an analysis of loss of heat removal accident in spent fuel pool of BWR 4 and a comparison of two different modelling approaches. The one-dimensional system thermal-hydraulic computer code RELAP5 and CFD tool ANSYS Fluent were used for the analysis. The results are similar, but the local effects cannot be simulated using a one-dimensional code. The ANSYS Fluent calculation demonstrated that this three-dimensional treatment allows to avoid the need for many one-dimensional modelling assumptions in the pool modelling and enables to reduce the uncertainties associated with natural circulation flow calculation.
Effects of the deviation characteristics of nuclear waste emplacement boreholes on borehole liner stresses; Yucca Mountain Project

Energy Technology Data Exchange (ETDEWEB)

Glowka, D.A.

1990-09-01

This report investigates the effects of borehole deviation on the useability of lined boreholes for the disposal of high-level nuclear waste at the proposed Yucca Mountain Repository in Nevada. Items that lead to constraints on borehole deviation include excessive stresses that could cause liner failure and possible binding of a waste container inside the liner during waste emplacement and retrieval operations. Liner stress models are developed for two general borehole configurations, one for boreholes drilled with a steerable bit and one for boreholes drilled with a non-steerable bit. Procedures are developed for calculating liner stresses that arise both during insertion of the liner into a borehole and during the thermal expansion process that follows waste emplacement. The effects of borehole curvature on the ability of the waste container to pass freely inside the liner without binding are also examined. Based on the results, specifications on borehole deviation allowances are developed for specific vertical and horizontal borehole configurations of current interest. 11 refs., 22 figs., 4 tabs.
Reactor-specific spent fuel discharge projections: 1985 to 2020

International Nuclear Information System (INIS)

Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

1986-09-01

The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel
Conditioning experience for spent radium sources

International Nuclear Information System (INIS)

Kang, I. S.; Shon, J. S.; Kim, K. J.; Min, D. K.

2001-01-01

In order to avoid accidents that could be resulted from improper storage of spent radium sources, it is necessary to condition and store them safely. The program for safe conditioning of spent radium sources by IAEA has been established to assist the developing countries. The main object of this paper is to apply the technology that was adapted by IAEA for the conditioning the national inventory of Ra-226 sources in member states, as a part of IAEA's project with the Korean expert team. This paper is the result that the Korean expert team carried out spent radium conditioning, under the project title 'Radium Conditioning in Myanmar(INT4131-06646C)'. The whole inventory of spent radium sources 1,429.5mCi, was safely conditioned by the Korean expert team according to the manual under the supervision of IAEA's technical officer and the control of Myanmar authority on behalf of Myanmar. These sources were encapsuled and welded into 27 small capsules and 3 large capsules, and conditioned in 3 lead shields, producing 3 concrete-shielded drums. The inventories were distributed into 3 shielding devices, holding 500mCi, 459.5mCi, and 470mCi
Lower Crstal Reflectity bands and Magma Emplacement in Norweigian sea, NE Atlantic

Science.gov (United States)

Rai, A.; Breivik, A. J.; Mjelde, R.

2013-12-01

In this study we present the OBS data collected along seismic profiles in the norweigian sea. The traveltime modelling of the OBS data provides first-hand information about seismic structure of the subsurface. However, waveform modelling is used to further constrain the fine scale structure, velocity constrast and velocity gradients. By forward modelling and inversion of the seismic waveforms, we show that the multiple bands of reflectivity could be due to multiple episodes of magma emplacements that might have frozen in the form of sills. These mafic intrusions probably intruded into the ductile lower crust during the main rifting phase of Europe and Greenland.
Lack of inhibiting effect of oil emplacement on quartz cementation: Evidence from Cambrian reservoir sandstones, Paleozoic Baltic Basin

DEFF Research Database (Denmark)

Molenaar, Nicolaas; Cyziene, Jolanta; Sliaupa, Saulius

2008-01-01

Currently, the question of whether or not the presence of oil in sandstone inhibits quartz cementation and preserves porosity is still debated. Data from a number of Cambrian sandstone oil fields and dry fields have been studied to determine the effects of oil emplacement on quartz cementation. T...
Transportation capabilities study of DOE-owned spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1994-10-01

This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.
Worldwide spent fuel transportation logistics

International Nuclear Information System (INIS)

Best, R.E.; Garrison, R.F.

1978-01-01

This paper presents an overview of the worldwide transportation requirements for spent fuel. Included are estimates of numbers and types of shipments by mode and cask type for 1985 and the year 2000. In addition, projected capital and transportation costs are presented. For the year 1977 and prior years inclusive, there is a cumulative worldwide requirement for approximately 300 MTU of spent fuel storage at away-from-reactor (AFR) facilities. The cumulative requirements for years through 1985 are projected to be nearly 10,000 MTU, and for the years through 2000 the requirements are conservatively expected to exceed 60,000 MTU. These AFR requirements may be related directly to spent fuel transportation requirements. In total nearly 77,000 total cask shipments of spent fuel will be required between 1977 and 2000. These shipments will include truck, rail, and intermodal moves with many ocean and coastal water shipments. A limited number of shipments by air may also occur. The US fraction of these is expected to include 39,000 truck shipments and 14,000 rail shipments. European shipments to regional facilities are expected to be primarily by rail or water mode and are projected to account for 16,000 moves. Pacific basin shipments will account for 4500 moves. The remaining are from other regions. Over 400 casks will be needed to meet the transportation demands. Capital investment is expected to reach $800,000,000 in 1977 dollars. Cumulative transport costs will be a staggering $4.4 billion dollars
US spent fuel research and experience

Energy Technology Data Exchange (ETDEWEB)

Machiels, A [EPRI and USDOE (United States)

2012-07-01

The structural performance of high-burnup spent fuel cladding during dry storage and transportation has been the subject of research and evaluation at EPRI for several years. The major issues addressed in this research program have included the following: Characterization and development of predictive models for damage mechanisms perceived to be potentially active during dry storage; Modeling and analysis of deformation processes during long-term dry storage; Development of cladding failure models and failure criteria, considering cladding material and physical conditions during dry storage and transportation; Failure analysis, considering end-of-dry-storage conditions, of spent fuel systems subjected to normal and accident conditions of transport, prescribed in Part 71 of Title 10 of the Code of Federal Regulations (10CFR71) While issues related to dry storage have largely been resolved, transportation issues have not, at least for spent fuel with discharge burnups greater than 45 GWd/MTU. A research program was launched in late 2002 following two NRC-industry meetings held on September 6, 2002 and October 23, 2002. The aim of the research program was to assess the performance of high-burnup spent fuel cladding under normal and accident conditions of transportation, as prescribed by 10CFR71, considering the physical characteristics and mechanical properties of cladding at the end of dry storage. The objective is to present a synthesis of the information that collectively forms a part of a technical basis intended to facilitate resolution of regulatory issues associated with the transportation of spent nuclear fuel characterized by discharge burnups greater than 45 GWd/MTU.
Spent Nuclear Fuel Project dose management plan

International Nuclear Information System (INIS)

Bergsman, K.H.

1996-03-01

This dose management plan facilitates meeting the dose management and ALARA requirements applicable to the design activities of the Spent Nuclear Fuel Project, and establishes consistency of information used by multiple subprojects in ALARA evaluations. The method for meeting the ALARA requirements applicable to facility designs involves two components. The first is each Spent Nuclear Fuel Project subproject incorporating ALARA principles, ALARA design optimizations, and ALARA design reviews throughout the design of facilities and equipment. The second component is the Spent Nuclear Fuel Project management providing overall dose management guidance to the subprojects and oversight of the subproject dose management efforts
Prospects of spent management in Spain

International Nuclear Information System (INIS)

Melches, C.; Ramirez, E.; Selgas, F.; Cabanilles, P.A.; Lopez Perez, B.; Uriarte, A.

1978-01-01

The purpose of this paper is to outline the forecast on spent fuel management in Spain, taking into account the international developments produced during the last years and specially on LWR fuels. This forecast is based on the following actions: increase of the storage capacity in the reactors: construction of an independent spent fuel storage installation (ISFSI) and a fuel reprocessing pilot plant. (author)
Time Well Spent? Relating Television Use to Children’s Free-Time Activities

Science.gov (United States)

Vandewater, Elizabeth A.; Bickham, David S.; Lee, June H.

2010-01-01

OBJECTIVES This study assessed the claim that children’s television use interferes with time spent in more developmentally appropriate activities. METHODS Data came from the first wave of the Child Development Supplement, a nationally representative sample of children aged 0 to 12 in 1997 (N = 1712). Twenty-four-hour time-use diaries from 1 randomly chosen weekday and 1 randomly chosen weekend day were used to assess children’s time spent watching television, time spent with parents, time spent with siblings, time spent reading (or being read to), time spent doing homework, time spent in creative play, and time spent in active play. Ordinary least squares multiple regression was used to assess the relationship between children’s television use and time spent pursuing other activities. RESULTS Results indicated that time spent watching television both with and without parents or siblings was negatively related to time spent with parents or siblings, respectively, in other activities. Television viewing also was negatively related to time spent doing homework for 7- to 12-year-olds and negatively related to creative play, especially among very young children (younger than 5 years). There was no relationship between time spent watching television and time spent reading (or being read to) or to time spent in active play. CONCLUSIONS The results of this study are among the first to provide empirical support for the assumptions made by the American Academy of Pediatrics in their screen time recommendations. Time spent viewing television both with and without parents and siblings present was strongly negatively related to time spent interacting with parents or siblings. Television viewing was associated with decreased homework time and decreased time in creative play. Conversely, there was no support for the widespread belief that television interferes with time spent reading or in active play. PMID:16452327
5 CFR 551.425 - Time spent receiving medical attention.

Science.gov (United States)

2010-01-01

... 5 Administrative Personnel 1 2010-01-01 2010-01-01 false Time spent receiving medical attention... Relation to Other Activities § 551.425 Time spent receiving medical attention. (a) Time spent waiting for and receiving medical attention for illness or injury shall be considered hours of work if: (1) The...

Development of on-site spent fuel transfer system designs

International Nuclear Information System (INIS)

Lambert, R.W.; Pennington, C.W.; Guerra, G.V.

1993-01-01

The Electric Power Research Institute (EPRI) of the United States has sponsored development of conceptual designs for accomplishing spent fuel transfer from spent fuel pools to casks and from one cask to another. Under an EPRI research contract, transnuclear has developed several concepts for spent fuel transfer systems. (J.P.N.)
Feasibility and incentives for the consideration of spent fuel operating histories in the criticality analysis of spent fuel shipping casks

International Nuclear Information System (INIS)

Sanders, T.L.; Westfall, R.M.; Jones, R.H.

1987-08-01

Analyses have been completed that indicate the consideration of spent fuel histories (''burnup credit'') in the design of spent fuel shipping casks is a justifiable concept that would result in cost savings and public risk benefits in the transport of spent nuclear fuel. Since cask capacities could be increased over those of casks without burnup credit, the number of shipments necessary to transport a given amount of fuel could be reduced. Reducing the number of shipments would increase safety benefits by reducing public and occupational exposure to both radiological and nonradiological risks associated with the transport of spent fuel. Economic benefits would include lower in-transit shipping, reduced transportation fleet capital costs, and reduced numbers of cask handling operations at both shipping and receiving facilities. 44 refs., 66 figs., 28 tabs
Studies on recycling and utilization of spent catalysts. Preparation of active hydrodemetallization catalyst compositions from spent residue hydroprocessing catalysts

Energy Technology Data Exchange (ETDEWEB)

Marafi, Meena; Stanislaus, Antony [Petroleum Refining Department, Petroleum Research and Studies Center, Kuwait Institute for Scientific Research, P.O. Box 24885, Safat (Kuwait)

2007-02-15

Spent catalysts form a major source of solid wastes in the petroleum refining industries. Due to environmental concerns, increasing emphasis has been placed on the development of recycling processes for the waste catalyst materials as much as possible. In the present study the potential reuse of spent catalysts in the preparation of active new catalysts for residual oil hydrotreating was examined. A series of catalysts were prepared by mixing and extruding spent residue hydroprocessing catalysts that contained C, V, Mo, Ni and Al{sub 2}O{sub 3} with boehmite in different proportions. All prepared catalysts were characterized by chemical analysis and by surface area, pore volume, pore size and crushing strength measurements. The hydrodesulfurization (HDS) and hydrodemetallization (HDM) activities of the catalysts were evaluated by testing in a high pressure fixed-bed microreactor unit using Kuwait atmospheric residue as feed. A commercial HDM catalyst was also tested under similar operating conditions and their HDS and HDM activities were compared with that of the prepared catalysts. The results revealed that catalyst prepared with addition of up to 40 wt% spent catalyst to boehmite had fairly high surface area and pore volume together with large pores. The catalyst prepared by mixing and extruding about 40 wt% spent catalyst with boehmite was relatively more active for promoting HDM and HDS reactions than a reference commercial HDM catalyst. The formation of some kind of new active sites from the metals (V, Mo and Ni) present in the spent catalyst is suggested to be responsible for the high HDM activity of the prepared catalyst. (author)
Spent Fuel Storage Operation - Lessons Learned

International Nuclear Information System (INIS)

2013-12-01

Experience gained in planning, constructing, licensing, operating, managing and modifying spent fuel storage facilities in some Member States now exceeds 50 years. Continual improvement is only achieved through post-project review and ongoing evaluation of operations and processes. This publication is aimed at collating and sharing lessons learned. Hopefully, the information provided will assist Member States that already have a developed storage capability and also those considering development of a spent nuclear fuel storage capability in making informed decisions when managing their spent nuclear fuel. This publication is expected to complement the ongoing Coordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR-III); the scope of which prioritizes facility operational practices in lieu of fuel and structural components behaviour over extended durations. The origins of the current publication stem from a consultants meeting held on 10-12 December 2007 in Vienna, with three participants from the IAEA, Slovenia and USA, where an initial questionnaire on spent fuel storage was formulated (Annex I). The resultant questionnaire was circulated to participants of a technical meeting, Spent Fuel Storage Operations - Lessons Learned. The technical meeting was held in Vienna on 13-16 October 2008, and sixteen participants from ten countries attended. A consultants meeting took place on 18-20 May 2009 in Vienna, with five participants from the IAEA, Slovenia, UK and USA. The participants reviewed the completed questionnaires and produced an initial draft of this publication. A third consultants meeting took place on 9-11 March 2010, which six participants from Canada, Hungary, IAEA, Slovenia and the USA attended. The meeting formulated a second questionnaire (Annex II) as a mechanism for gaining further input for this publication. A final consultants meeting was arranged on 20-22 June 2011 in Vienna. Six participants from Hungary, IAEA, Japan
Spent fuel storage requirements, 1991--2040

International Nuclear Information System (INIS)

1991-12-01

Historical inventories of spent fuel are combined with US Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements over the next 50 years, through the year 2040. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. Historical data through December 1990 are derived from the 1991 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges
Spent Nuclear Fuel (SNF) Project Execution Plan

International Nuclear Information System (INIS)

LEROY, P.G.

2000-01-01

The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities
Spent Nuclear Fuel (SNF) Project Execution Plan

Energy Technology Data Exchange (ETDEWEB)

LEROY, P.G.

2000-11-03

The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.
Safeguardability of advanced spent fuel conditioning process

Energy Technology Data Exchange (ETDEWEB)

Li, T. K. (Tien K.); Lee, S. Y. (Sang Yoon); Burr, Tom; Russo, P. A. (Phyllis A.); Menlove, Howard O.; Kim, H. D.; Ko, W. I. (Won Il); Park, S. W.; Park, H. S.

2004-01-01

The Advanced Spent Fuel Conditioning Process (ACP) is an electro-metallurgical treatment technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology since 1977 for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. By using of this technology, a significant reduction of the volume and heat load of spent fuel is expected, which would lighten the burden of final disposal in terms of disposal size, safety and economics. In the framework of collaboration agreement to develop the safeguards system for the ACP, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and the KAERI since 2002. In this study, the safeguardability of the ACP technology was examined for the pilot-scale facility. The process and material flows were conceptually designed, and the uncertainties in material accounting were estimated with international target values.
Electrodialytic decontamination of spent ion exchange resins

International Nuclear Information System (INIS)

Nott, B.R.

1982-01-01

Development of a novel electrodialytic decontamination process for the selective removal of radioactive Cs from spent ion exchange resins containing large amounts of Li is described. The process involves passage of a dc electric current through a bed of the spent ion exchange resin in a specially designed electrodialytic cell. The radiocesium so removed from a volume of the spent resin is concentrated onto a much smaller volume of a Cs selective sorbent to achieve a significant radioactive waste volume reduction. Technical feasibility of the electrodialytic resin decontamination process has been demonstrated on a bench scale with a batch of simulated spent ion exchange resin and using potassium cobalt ferrocyanide as the Cs selective sorbent. A volume reduction factor between 10 and 17 has been estimated. The process appears to be economically attractive. Improvements in process economics can be expected from optimization of the process. Other possible applications of the EDRD process have been identified
Spent fuel management of NPPs in Argentina

International Nuclear Information System (INIS)

Alvarez, D.E.; Lee Gonzalez, H.M.

2010-01-01

There are two Nuclear Power Plants in operation in Argentina: 'Atucha I' (unique PHWR design) in operation since 1974, and 'Embalse' (typical Candu reactor) which started operation in 1984. Both NPPs are operated by 'Nucleoelectrica Argentina S.A' which is responsible for the management and interim storage of spent fuel till the end of the operative life of the plants. A third NPP, 'Atucha II' is under construction, with a similar design of Atucha I. The legislative framework establishes that after final shutdown of a NPP the spent fuel will be transferred to the 'National Atomic Energy Commission', which is also responsible for the decommissioning of the Plants. In Atucha I, the spent fuel is stored underwater, until another option is implemented meanwhile in Embalse the spent fuel is stored during six years in pools and then it is moved to a dry storage. A decision about the fuel cycle back-end strategy will be taken before year 2030. (authors)
Landsat investigations of the northern Paradox basin, Utah and Colorado: implications for radioactive waste emplacement

Science.gov (United States)

Friedman, Jules D.; Simpson, Shirley L.

1978-01-01

The first stages of a remote-sensing project on the Paradox basin, part of the USGS (U.S. Geological Survey) radioactive waste-emplacement program, consisted of a review and selection of the best available satellite scanner images to use in geomorphologic and tectonic investigations of the region. High-quality Landsat images in several spectral bands (E-2260-17124 and E-5165-17030), taken under low sun angle October 9 and 10, 1975, were processed via computer for planimetric rectification, histogram analysis, linear transformation of radiance values, and edge enhancement. A lineament map of the northern Paradox basin was subsequently compiled at 1:400,000 using the enhanced Landsat base. Numerous previously unmapped northeast-trending lineaments between the Green River and Yellowcat dome; confirmatory detail on the structural control of major segments of the Colorado, Gunnison, and Dolores Rivers; and new evidence for late Phanerozoic reactivation of Precambrian basement structures are among the new contributions to the tectonics of the region. Lineament trends appear to be compatible with the postulated Colorado lineament zone, with geophysical potential-field anomalies, and with a northeast-trending basement fault pattern. Combined Landsat, geologic, and geophysical field evidence for this interpretation includes the sinuousity of the composite Salt Valley anticline, the transection of the Moab-Spanish Valley anticline on its southeastern end by northeast-striking faults, and possible transection (?) of the Moab diapir. Similarly, northeast-trending lineaments in Cottonwood Canyon and elsewhere are interpreted as manifestations of structures associated with northeasterly trends in the magnetic and gravity fields of the La Sal Mountains region. Other long northwesterly lineaments near the western termination of the Ryan Creek fault zone. may be associated with the fault zone separating the Uncompahgre horst uplift from the Paradox basin. Implications of the
Spent sulfite liquor developments

Energy Technology Data Exchange (ETDEWEB)

Black, H H

1958-01-01

A review of methods of utilizing spent sulfite liquor, including evaporation and burning, fermentation to produce yeast or alcohol, production of vanillin and lignosulfonates, and use as a roadbinder.
Anisotropy of magnetic susceptibility of silicic rocks from quarries in the vicinity of São Marcos, Rio Grande do Sul, South Brazil: Implications for emplacement mechanisms

Science.gov (United States)

Cañón-Tapia, Edgardo; Raposo, M. Irene B.

2018-04-01

The Paraná-Etendeka Large Igneous Province includes acid volcanic rocks that can be found throughout its extension. Several aspects concerning those rocks remain controversial, including their mechanism of emplacement and location of their eruptive sources. Opening of several quarries of dimension stone near the city of Sao Marcos, Rio Grande do Sul, Brazil, offers a unique opportunity to study in detail the acid products. Here, we present the results of a study of the anisotropy of magnetic susceptibility (AMS) completed in some rocks that had been interpreted as the roots of volcanic conduits. Our results, and reexamination of the textural features of the rocks, lead to a reinterpretation that suggests that these rocks were emplaced subaerially, and involved assimilation and remelting of clastic components of previous products. Due to the inferred conditions of emplacement, it is unlikely that the eruptive vents are located far from the area of study, therefore ruling out the long-travelled nature of these products.
Spent fuel management and closed nuclear fuel cycle

International Nuclear Information System (INIS)

Kudryavtsev, E.G.

2012-01-01

Strategic objectives set by Rosatom Corporation in the field of spent fuel management are given. By 2030, Russia is to create technological infrastructure for innovative nuclear energy development, including complete closure of the nuclear fuel cycle. A target model of the spent NPP nuclear fuel management system until 2030 is analyzed. The schedule for key stages of putting in place the infrastructure for spent NPP fuel management is given. The financial aspect of the problem is also discussed [ru
Electrochemical processing of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Williamson, M. A.; Willit, J. L.; Barnes, L. A.; Figueroa, J.; Limmer, S. L.; Blaskovitz, R. [Argonne National Laboratory, Argonne (United States)

2008-08-15

Our work in developing the fuel cycles and electrochemical technologies needed for the treatment of spent light water reactor and spent fast reactor fuel is progressing well. Baseline flowsheets along with a theoretical material balance have been developed for treatment of each type of fuel. A discussion about the flowsheets provides the opportunity to present the status of our technology development activities and future research and development directions.
Electrochemical processing of spent nuclear fuel

International Nuclear Information System (INIS)

Williamson, M. A.; Willit, J. L.; Barnes, L. A.; Figueroa, J.; Limmer, S. L.; Blaskovitz, R.

2008-01-01

Our work in developing the fuel cycles and electrochemical technologies needed for the treatment of spent light water reactor and spent fast reactor fuel is progressing well. Baseline flowsheets along with a theoretical material balance have been developed for treatment of each type of fuel. A discussion about the flowsheets provides the opportunity to present the status of our technology development activities and future research and development directions
Reactor-specific spent fuel discharge projections: 1986 to 2020

International Nuclear Information System (INIS)

Heeb, C.M.; Walling, R.C.; Purcell, W.L.

1987-03-01

The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No new orders with extended burnup, (2) No new orders with constant burnup, (3) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel. 6 refs., 8 figs., 8 tabs
Pyroprocessing oxide spent nuclear fuels for efficient disposal

International Nuclear Information System (INIS)

McPheeters, C.C.; Pierce, R.D.; Mulcahey, T.P.

1994-01-01

Pyrochemical processing as a means for conditioning spent nuclear fuels for disposal offers significant advantages over the direct disposal option. The advantages include reduction in high-level waste volume; conversion of most of the high-level waste to a low-level waste in which nearly all the transuranics (TRU) have been removed; and incorporation of the TRUs into a stable, highly radioactive waste form suitable for interim storage, ultimate destruction, or repository disposal. The lithium process has been under development at Argonne National Laboratory for use in pyrochemical conditioning of spent fuel for disposal. All of the process steps have been demonstrated in small-scale (0.5-kg simulated spent fuel) experiments. Engineering-scale (20-kg simulated spent fuel) demonstration of the process is underway, and small-scale experiments have been conducted with actual spent fuel from a light water reactor (LWR). The lithium process is simple, operates at relatively low temperatures, and can achieve high decontamination factors for the TRU elements. Ordinary materials, such as carbon steel, can be used for process containment
BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

Energy Technology Data Exchange (ETDEWEB)

Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-10-01

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is
Immobilization of spent resin with epoxy resin

International Nuclear Information System (INIS)

Gultom, O.; Suryanto; Sayogo; Ramdan

1997-01-01

immobilization of spent resin using epoxy resin has been conducted. The spent resin was mixtured with epoxy resin in variation of concentration, i.e., 30, 40, 50, 60, 70 weight percent of spent resin. The mixture were pour into the plastic tube, with a diameter of 40 mm and height of 40 mm. The density, compressive strength and leaching rate were respectively measured by quanta chrome, paul weber apparatus and gamma spectrometer. The results showed that the increasing of waste concentration would be decreased the compressive strength, and increased density by immobilized waste. The leaching rate of 137 Cs from waste product was not detected in experiment (author)

Radioactive waste isolation in salt: Peer review of the Fluor Technology, Inc., report and position paper concerning waste emplacement mode and its effect on repository conceptual design

International Nuclear Information System (INIS)

Hambley, D.F.; Russell, J.E.; Whitfield, R.G.

1987-02-01

Recommendations for revising the Fluor Technology, Inc., draft position paper entitled Evaluation of Waste Emplacement Mode and the final report entitled Waste Package/Repository Impact Study include: reevaluate the relative rankings for the various emplacement modes; delete the following want objectives: maximize ability to locate the package horizon because sufficient flexibility exists to locate rooms in the relatively clean San Andres Unit 4 Salt and maximize far-field geologic integrity during retrieval because by definition the far field will be unaffected by thermal and stress perturbations caused by remining; give greater emphasis to want objectives regarding cost and use of present technology; delete the following statements from pages 1-1 and 1-2 of the draft position paper: ''No thought or study was given to the impacts of this configuration [vertical emplacement] on repository construction or short and long-term performance of the site'' and ''Subsequent salt repository designs adopted the vertical emplacement configuration as the accepted method without further evaluation.''; delete App. E and lines 8-17 of page 1-4 of the draft position paper because they are inappropriate; adopt a formal decision-analysis procedure for the 17 identified emplacement modes; revise App. F of the impact study to more accurately reflect current technology; consider designing the underground layout to take advantage of stress-relief techniques; consider eliminating reference to fuel assemblies <10 yr ''out-of-reactor''; model the temperature distribution, assuming that the repository is constructed in an infinitely large salt body; state that the results of creep analyses must be considered tentative until they can be validated by in situ measurements; and reevaluate the peak radial stresses on the waste package so that the calculated stress conditions more closely approximate expected in situ conditions
Dry refabrication technology development of spent nuclear fuel

International Nuclear Information System (INIS)

Park, Geun Il; Lee, J. W.; Song, K. C.

2012-04-01

Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed
Stand for visual ultrasonic testing of spent fuel

International Nuclear Information System (INIS)

Czajkowski, W.; Borek-Kruszewska, E.

2001-01-01

A stand for visual and ultrasonic testing of spent fuel, constructed under Strategic Governmental Programme for management of spent fuel and radioactive waste, is presented in the paper. The stand, named 'STEND-1', built up at the Institute of Atomic Energy in Swjerk, is appointed for underwater visual testing of spent fuel elements type MR6 and WWR by means of TV-CCD camera and image processing system and for ultrasonic scanning of external surface of these elements by means of video scan immersion transducer and straight UHT connector. 'STEND-1' is built using flexible in use, high-tensile, anodized aluminum profiles. All the profiles feature longitudinal grooves to accommodate connecting elements and for the attachment of accessories at any position. They are also characterised by straight-through core bores for use with standard fastening elements and to accommodate accessory components. Stand, equipped with automatic control and processing system based on personal computer, may be manually or automatically controlled. Control system of movements of the camera in the vertical axis and rotational movement of spent fuel element permits to fix chosen location of fuel element with accuracy better than 0.1 mm. High resolution of ultrasonic method allows to record damages of outer surface of order 0.1 mm. The results of visual testing of spent fuel are recorded on video tape and then may be stored on the hard disc of the personal computer and presented in shape of photo or picture. Only selected damage surfaces of spent fuel elements are tested by means of ultrasonic scanning. All possibilities of the stand and results of visual testing of spent fuel type WWR are presented in the paper. (author)
Multi-purpose container technologies for spent fuel management

International Nuclear Information System (INIS)

2000-12-01

The management of spent nuclear fuel is an integral part of the nuclear fuel cycle. Spent fuel management resides in the back end of the fuel cycle, and is not revenue producing as electric power generation is. It instead results in a cost associated power generation. It is a major consideration in the nuclear power industry today. Because technologies, needs and circumstances vary from country to country, there is no single, standardized approach to spent fuel management. The projected cumulative amount of spent fuel generated worldwide by 2010 will be 330 000 t HM. When reprocessing is accounted for, that amount is likely to be reduced to 215 000 t HM, which is still more than twice as much as the amount now in storage. Considering the limited capacity of at-reactor (AR) storage, various technologies are being developed for increasing storage capacities. At present, many countries are developing away-from-reactor (AFR) storage in the form of pool storage or as dry storage. Further these AFR storage systems may be at-reactor sites or away-from-reactor sites (e.g. centrally located interim storage facilities, serving several reactors). The dry storage technologies being developed are varied and include vaults, horizontal concrete modules, concrete casks, and metal casks. The review of the interim storage plans of several countries indicates that the newest approaches being pursued for spent fuel management use dual-purpose and multi-purpose containers. These containers are envisaged to hold several spent fuel assemblies, and be part of the transport, storage, and possibly geological disposal systems of an integrated spent fuel management system
Dry refabrication technology development of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Park, Geun Il; Lee, J. W.; Song, K. C.; and others

2012-04-15

Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed.
Spent fuel surveillance and monitoring methods

International Nuclear Information System (INIS)

1988-05-01

The Technical Committee Meeting on ''Spent Fuel Surveillance and Monitoring Methods'' (27-30 October 1987) has been organized in accordance with recommendations of the International Standing Advisory Group on Spent Fuel Management during its second meeting in 1986. The aim of the meeting was to discuss the above questions with emphasis on current design and operation criteria, safety principles and licensing requirements and procedures in order to prevent: inadvertent criticality, undue radiation exposure, unacceptable release of radioactivity as well as control for loss of storage pool water, crud impact, water chemistry, distribution and behaviour of particulates in cooling water, oxidation of intact and failed fuel rods as a function of temperature and burnup; distribution of radiation and temperature through dry cask wall, monitoring of leakages from pools and gas escapes from dry storage facilities, periodical integrity tests of the containment barriers, responsibilities of organizations for the required operation, structure, staff and subordination, etc. The presentations of the Meeting were divided into two sessions: Spent fuel surveillance programmes and practice in Member States (4 papers); Experimental methods developed in support of spent fuel surveillance programmes (5 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs
Transportation and storage of foreign spent power reactor fuel

International Nuclear Information System (INIS)

1979-01-01

This report describes the generic actions to be taken by the Department of Energy, in cooperation with other US government agencies, foreign governments, and international organizations, in support of the implementation of Administration policies with respect to the following international spent fuel management activities: bilateral cooperation related to expansion of foreign national storage capacities; multilateral and international cooperation related to development of multinational and international spent fuel storage regimes; fee-based transfer of foreign spent power reactor fuel to the US for storage; and emergency transfer of foreign spent power reactor fuel to the US for storage
Preliminary concepts: safeguards for spent light-water reactor fuels

International Nuclear Information System (INIS)

Cobb, D.D.; Dayem, H.A.; Dietz, R.J.

1979-06-01

The technology available for safeguarding spent nuclear fuels from light-water power reactors is reviewed, and preliminary concepts for a spent-fuel safeguards system are presented. Essential elements of a spent-fuel safeguards system are infrequent on-site inspections, containment and surveillance systems to assure the integrity of stored fuel between inspections, and nondestructive measurements of the fuel assemblies. Key safeguards research and development activities necessary to implement such a system are identified. These activities include the development of tamper-indicating fuel-assembly identification systems and the design and development of nondestructive spent-fuel measurement systems
Spent fuel storage requirements, 1990--2040

International Nuclear Information System (INIS)

Walling, R.; Bierschbach, M.

1990-11-01

Historical inventories of spent fuel are combined with US Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements over the next 51 years, through the year 2040. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. Historical data through December 1989 are derived from the 1990 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 15 refs., 3 figs., 11 tabs
Loss of spent fuel pool cooling PRA: Model and results

International Nuclear Information System (INIS)

Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

1996-09-01

This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 x 10 -5 and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 x 10 -3 . Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible
Granite ascent and emplacement during contractional deformation in convergent orogens

Science.gov (United States)

Brown, Michael; Solar, Gary S.

1998-09-01

Based on a case study in the Central Maine Belt of west-central Maine, U.S.A., it is proposed that crustal-scale shear zone systems provide an effective focussing mechanism for transfer of granite melt through the crust in convergent orogens. During contractional deformation, flow of melt in crustal materials at depths below the brittle-plastic transition is coupled with plastic deformation of these materials. The flow is driven by pressure gradients generated by buoyancy forces and tectonic stresses. Within the oblique-reverse Central Maine Belt shear zone system, stromatic migmatite and concordant to weakly discordant irregular granite sheets occur in zones of higher strain, which suggests percolative flow of melt to form the migmatite leucosomes and viscous flow of melt channelized in sheet-like bodies, possibly along fractures. Cyclic fluctuations of melt pressure may cause instantaneous changes in the effective permeability of the flow network if self-propagating melt-filled tensile and/or dilatant shear fractures are produced due to melt-enhanced embrittlement. Inhomogeneous migmatite and schlieric granite occur in zones of lower strain, which suggests migration of partially-molten material through these zones en masse by granular flow, and channelized flow of melt carrying entrained residue. Founded on the Central Maine Belt case study, we develop a model of melt extraction and ascent using the driving forces, stress conditions and crustal rheologies in convergent, especially transpressive orogens. Ascent of melt becomes inhibited with decreasing depth as the solidus is approached. For intermediate a(H 2O) muscovite-dehydration melting, the water-saturated solidus occurs between 400 and 200 MPa, near the brittle-plastic transition during high- T-low- P metamorphism, where the balance of forces favors (sub-) horizontal fracture propagation. Emplacement of melt may be accommodated by ductile flow and/or stoping of wall rock, and inflation may be accommodated
Spent Fuel Transportation Package Performance Study - Experimental Design Challenges

International Nuclear Information System (INIS)

Snyder, A. M.; Murphy, A. J.; Sprung, J. L.; Ammerman, D. J.; Lopez, C.

2003-01-01

Numerous studies of spent nuclear fuel transportation accident risks have been performed since the late seventies that considered shipping container design and performance. Based in part on these studies, NRC has concluded that the level of protection provided by spent nuclear fuel transportation package designs under accident conditions is adequate. [1] Furthermore, actual spent nuclear fuel transport experience showcase a safety record that is exceptional and unparalleled when compared to other hazardous materials transportation shipments. There has never been a known or suspected release of the radioactive contents from an NRC-certified spent nuclear fuel cask as a result of a transportation accident. In 1999 the United States Nuclear Regulatory Commission (NRC) initiated a study, the Package Performance Study, to demonstrate the performance of spent fuel and spent fuel packages during severe transportation accidents. NRC is not studying or testing its current regulations, a s the rigorous regulatory accident conditions specified in 10 CFR Part 71 are adequate to ensure safe packaging and use. As part of this study, NRC currently plans on using detailed modeling followed by experimental testing to increase public confidence in the safety of spent nuclear fuel shipments. One of the aspects of this confirmatory research study is the commitment to solicit and consider public comment during the scoping phase and experimental design planning phase of this research
Spent fuel management in Spain

International Nuclear Information System (INIS)

Gago, J.A.; Gravalos, J.M.

1996-01-01

There are presently nine Light Water Reactors in operation, representing around a 34% of the overall electricity production. In the early years, a small amount of spent fuel was sent to be reprocessed, although this policy was cancelled in favor of the open cycle option. A state owned company, ENRESA, was created in 1984, which was given the mandate to manage all kinds of radioactive wastes generated in the country. Under the present scenario, a rough overall amount of 7000 tU of spent fuel will be produced during the lifetime of the plants, which will go into final disposal. (author)
Repository thermal response: A preliminary evaluation of the effects of modeled waste stream resolution

International Nuclear Information System (INIS)

Ryder, E.E.; Dunn, E.

1995-09-01

One of the primary factors that influences our predictions of host-rock thermal response within a high level waste repository is how the waste stream's represented in the models. In the context of thermal modeling, waste stream refers to an itemized listing of the type (pressurized-water or boiling-water reactor), age, burnup, and enrichment of the spent nuclear fuel assemblies entering the repository over the 25-year emplacement phase. The effect of package-by-package variations in spent fuel characteristics on predicted repository thermal response is the focus of this report. A three-year portion of the emplacement period was modeled using three approaches to waste stream resolution. The first assumes that each package type emplaced in a given year is adequately represented by average characteristics. For comparison, two models that explicitly account for each waste package's individual characteristics were run; the first assuming a random selection of packages and the second an ordered approach aimed at locating the higher power output packages toward the center of the emplacement area. Results indicate that the explicit representation of packages results in hot and cold spots that could have performance assessment and design implications. Furthermore, questions are raised regarding the representativeness of average characteristics with respect to integrated energy output and the possible implications of a mass-based repository loading approach
Transitions in Lava Emplacement Recorded in the Deccan Traps Sequence (India)

Science.gov (United States)

Vanderkluysen, L.; Self, S.; Jay, A. E.; Sheth, H. C.; Clarke, A. B.

2015-12-01

Transitions in the style of lava flow emplacement are recognized in the stratigraphic sequence of several mafic large igneous provinces (LIPs), including the Etendeka (Namibia), the Faeroe Islands (North Atlantic LIP), the Ethiopian Traps, and the Deccan Traps (India). These transitions, from units dominated by meter-sized pāhoehoe toes and lobes to those dominated by inflated sheet lobes tens to hundreds of meters in width and meters to tens of meters in height, seems to be a fundamental feature of LIP emplacement. In the Deccan, this volcanological transition is thought to coincide with deeper changes to the volcano-magmatic system expressed, notably, in the trace element and isotopic signature of erupted flows. We investigated this transition in the Deccan Traps by logging eight sequences along the Western Ghats, an escarpment in western India where the Deccan province is thickest and best exposed. The Deccan province, which once covered ~1 million km2 of west-central India, is subdivided in eleven chemo-stratigraphic formations in the type sections of the Western Ghats. Where the lower Deccan formations are exposed, we found that as much as 65% of the exposed thickness (below the Khandala Formation) is made up of sheet lobes, from 40% in the Bhimashankar Formation to 75% in the Thakurvadi Formation. Near the bottom of the sequence, 25% of the Neral Formation is composed of sheet lobes ≥15 m in thickness. On this basis, the traditional view that inflated sheet lobes are an exclusive feature of the upper part of the stratigraphy must be challenged. Several mechanisms have been proposed to explain the development of compound flows and inflated sheet lobes, involving one or more of the following factors: underlying slope, varying effusion rate, and source geometry. Analogue experiments are currently under way to test the relative influence of each of these factors in the development of different lava flow morphologies in LIPs.
Comparison of spent fuel management fee collection alternatives

International Nuclear Information System (INIS)

White, M.K.; Engel, R.L.

1979-01-01

Five alternative methods for recovering the costs of spent fuel management were evaluated. These alternatives consist of collecting the fee for various components of spent fuel management cost (AFR basin storage, transportation from AFR basin to the repository, packaging, repository, R and D, and government overhead) at times ranging from generation of power to delivery of the spent fuel to the government. The five fee collection mechanisms were analyzed to determine how well they serve the interests of the public and the electricity ratepayer
Spent fuel management: Current status and prospects

International Nuclear Information System (INIS)

1988-12-01

The main objective of the Advisory Group on Spent Fuel Management is to review the world-wide situation in Spent Fuel Management, to define the most important directions of national efforts and international cooperation in this area, to exchange information on the present status and progress in performing the back-end of Nuclear Fuel Cycle and to elaborate the general recommendations for future Agency programmes in the field of spent fuel management. This report which is a result of the third IAEA Advisory Group Meeting (the first and second were held in 1984 and 1986) is intended to provide the reader with an overview of the status of spent fuel management programmes in a number of leading countries, with a description of the past and present IAEA activities in this field of Nuclear Fuel Cycle and with the Agency's plans for the next years, based on the proposals and recommendations of Member States. A separate abstract was prepared for each of 14 papers presented at the advisory group meeting. Refs, figs and tabs
Corrosion of spent Advanced Test Reactor fuel

International Nuclear Information System (INIS)

Lundberg, L.B.; Croson, M.L.

1994-01-01

The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented
Technical Development on Burn-up Credit for Spent LWR Fuel

Energy Technology Data Exchange (ETDEWEB)

Gauld, I.C.

2001-12-26

Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.
Technical development on burn-up credit for spent LWR fuels

International Nuclear Information System (INIS)

Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori

2000-10-01

Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

Technical development on burn-up credit for spent LWR fuels

Energy Technology Data Exchange (ETDEWEB)

Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-10-01

Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)
Spent Nuclear Fuel Alternative Technology Decision Analysis

International Nuclear Information System (INIS)

Shedrow, C.B.

1999-01-01

The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology
Spent grain as energy source for breweries

Energy Technology Data Exchange (ETDEWEB)

Scharf, P; Meyer-Pittroff, R [Technical Univ. of Munich, Freising (West Germany)

1990-01-01

About 1.7 million tons of spent grain are produced from brewing operations each year in the Federal Republic of Germany. Because there is more spent grain than can be sold as cattle fodder, other uses for this product are being sought. These include composting, methane generation, and direct combustion. Researchers are using anaerobic fermentation as a means of producing biogas and using waste energy from the brewery to heat the fermenter. Combustion of spent grain can give a net energy of 12.8 MJ per kg dry grain. With biogas production, about 15-20% of the whole primary energy or 50% of the primary energy used in the boiling house could be substituted by the biogas.
Spent fuel storage and transportation - ANSTO experience

International Nuclear Information System (INIS)

Irwin, Tony

2002-01-01

The Australian Nuclear Science and Technology Organisation (ANSTO) has operated the 10 MW DIDO class High Flux Materials Test Reactor (HIFAR) since 1958. Refuelling the reactor produces about 38 spent fuel elements each year. Australia has no power reactors and only one operating research reactor so that a reprocessing plant in Australia is not an economic proposition. The HEU fuel for HIFAR is manufactured at Dounreay using UK or US origin enriched uranium. Spent fuel was originally sent to Dounreay, UK for reprocessing but this plant was shutdown in 1998. ANSTO participates in the US Foreign Research Reactor Spent Fuel Return program and also has a contract with COGEMA for the reprocessing of non-US origin fuel
Existing and near future practices of spent fuel storage in Slovak Republic

International Nuclear Information System (INIS)

Mizov, J.

1999-01-01

In this paper existing and near future practices of spent fuel storage in Slovak Republic are discussed: (1) Reactor operation and spent fuel production; (2) Past policy in spent fuel storage; (3) Away-from-reactor (AFR) storage facility at Bohunice NPP site; (4) Present policy in spent fuel storage; (5) Final disposal of spent fuel
Swedish nuclear waste efforts

International Nuclear Information System (INIS)

Rydberg, J.

1981-09-01

After the introduction of a law prohibiting the start-up of any new nuclear power plant until the utility had shown that the waste produced by the plant could be taken care of in an absolutely safe way, the Swedish nuclear utilities in December 1976 embarked on the Nuclear Fuel Safety Project, which in November 1977 presented a first report, Handling of Spent Nuclear Fuel and Final Storage of Vitrified Waste (KBS-I), and in November 1978 a second report, Handling and Final Storage of Unreprocessed Spent Nuclear Fuel (KBS II). These summary reports were supported by 120 technical reports prepared by 450 experts. The project engaged 70 private and governmental institutions at a total cost of US $15 million. The KBS-I and KBS-II reports are summarized in this document, as are also continued waste research efforts carried out by KBS, SKBF, PRAV, ASEA and other Swedish organizations. The KBS reports describe all steps (except reprocessing) in handling chain from removal from a reactor of spent fuel elements until their radioactive waste products are finally disposed of, in canisters, in an underground granite depository. The KBS concept relies on engineered multibarrier systems in combination with final storage in thoroughly investigated stable geologic formations. This report also briefly describes other activities carried out by the nuclear industry, namely, the construction of a central storage facility for spent fuel elements (to be in operation by 1985), a repository for reactor waste (to be in operation by 1988), and an intermediate storage facility for vitrified high-level waste (to be in operation by 1990). The R and D activities are updated to September 1981
DOE not planning to accept spent nuclear fuel

International Nuclear Information System (INIS)

Anon.

1995-01-01

Samuel K. Skinner, president of Commonwealth Edison Co. (ComEd), said open-quotes The federal government has a clear responsibility to begin accepting spent nuclear fuel in 1988,close quotes citing the Nuclear Waste Policy Act of 1982 before the Senate Energy and Natural Resources Committee. Based in Chicago, ComEd operates 12 nuclear units, making it the nation's largest nuclear utility. open-quotes Since 1983, the consumers who use electricity produced at all nuclear power plants have been paying to fund federal management of spent nuclear fuel. Consumer payments and obligations, with interest, now total more than $10 billion. Electricity consumers have held up their side of the deal. The federal government must do the same,close quotes Skinner added. Skinner represented the Nuclear Energy Institute (NEI) before the committee. NEI is the Washington-based trade association of the nuclear energy industries. For more than 12 years, utility customers have been paying one-tenth of a cent per kWhr to fund a federal spent fuel management program under the Nuclear Waste Policy Act of 1982. Under this act, the federal government assumed responsibility for management of spent fuel from the nation's nuclear power plants. The U.S. Department of Energy (DOE) was assigned to manage the storage and disposal program. DOE committed to begin accepting spent fuel from nuclear power plants by January 31, 1988. DOE has spent almost $5 million studying a site in Nevada, but is about 12 years behind schedule and does not plan to accept spent fuel beginning in 1998. DOE has said a permanent storage site will not be ready until 2010. This poses a major problem for many of the nation's nuclear power plants which supply about 20% of the electricity in the US
Status of Away From Reactor spent fuel storage program

International Nuclear Information System (INIS)

King, F.D.

1979-07-01

The Away From Reactor (AFR) Spent Fuel Program that the US Department of Energy established in 1977 is intended to preclude the shutting down of commercial nuclear power reactors because of lack of storage space for spent fuel. Legislation now being considered by Congress includes plans to provide storage space for commercial spent fuel beginning in 1983. Utilities are being encouraged to provide as much storage space as possible in their existing storage facilities, but projections indicate that a significant amount of AFR storage will be required. The government is evaluating the use of both existing and new storage facilities to solve this forecasted storage problem for commercial spent fuel
Spent fuel heatup following loss of water during storage

International Nuclear Information System (INIS)

Benjamin, A.S.; McCloskey, D.J.

1978-01-01

Spent fuel assemblies from light water reactors are typically stored for one year or more in the reactor spent fuel pool and then transported for long-term storage at an off-site location. Because of the design, construction, and operation features of spent fuel storage pools, an accident that might drain most of the water from a pool is assessed as being extremely improbable. As a limiting case, however, a hypothetical incident involving instantaneous draining of all the water from a storage pool has been postulated, and the subsequent heatup of the spent fuel elements has been evaluated. The model is analyzed, and results are summarized
Management of spent fuel in Republic of Korea

International Nuclear Information System (INIS)

Pak, Hyun-Soo; Seo, In-Seok; Pak, Sang-Ki.

1989-01-01

At present in Republic of Korea, 8 PWR and 1 CANDU power plants are in operation or under construction, and the total capacity of power generation facilities has become 7.6 GWe. In addition, two PWRs of more than 900 MWe each are expected to be constructed by mid 1990s. More than 50 % of the electric power demand was supplied by nuclear power generation since 1987, but the spent fuel generated in nuclear power plants is stored in storage water tanks in respective reactor sites. The total capacity of spent fuel to be stored in the AR facilities of 9 nuclear power plants is about 2730 MTU, and the spent fuel released from these reactors since 1980 is about 810 MTU. The present capacity of AR storage pools seems to be used up by mid 1990s. According to the revised Atomic Energy Acts in May, 1986, the government is to take the responsibility of spent fuel management, and the policy of constructing the storage facilities outside reactor sites by the end of 1997 was established by the Atomic Energy Commission. The responsibility of the management of spent fuel that exceeds the present capacity of AR pools is to be taken by KEPCO, therefore the preliminary analysis of the feasible option on the extension of AR facilities and the comprehensive management plan for spent fuel placing emphasis on the research and development of away-from-reactor storage were decided. (Kako, I.)
Development of compaction technique for spent fuel skeletons

Energy Technology Data Exchange (ETDEWEB)

Yoon, Ji Sup; Kim, Young Hwan; Jung, Jae Hoo [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-12-01

To increase the utilization of uranium resources contained in the spent fuel, the spent fuel is reused. For this, the spent fuel is dismantled or spent fuel rod is extracted from the spent fuel assembly. When the rod is extracted, the remaining components of spent fuel assembly, so called a NFBC(Non-Fuel Bearing Components), should be compacted for the final disposal. To this end, several companies developed the NFBC compactors. German company, named as GNS has developed the direct compression devices of the NFBCs for the rod consolidation and installed it at the PKA(2) of pilot conditioning plant. B and W (Babcock and Wilcox) in USA adopted cutting method rather than the compression method and developed the special cutting devices of NFBC which can be applied underwater environment. In this study the characteristics of these two methods was investigated, in terms of fabrication cost of devices, maintainability in a high radioactive environment, required power and work volume for operation. Also, the optimal power source is selected by comparing the maximum power versus the work volume for operation. In addition to these, the reduction ratio of the bulk volume is obtained while varying the cutting length of the NFBC through a series of experiments. Based on the results of analysis and experiments, the cutting method after compression is selected as an optimal volume reduction method and its design specification is obtained. 8 refs., 62 figs., 32 tabs. (Author)
ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

Energy Technology Data Exchange (ETDEWEB)

Chang, Sunyoung; Jeong, Yon Hong; Han, Jae-Jun; Lee, Aeri; Hwang, Yong-Soo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

2015-10-15

The disposal facility of the spent nuclear fuel will be operated from 2051. This paper presents the ENVI code developed by GoldSim Software to simulate options for managing spent nuclear fuel (SNF) in South Korea. The ENVI is a simulator to allow decision-makers to assist to evaluate the performance for spent nuclear fuel management. The multiple options for managing the spent nuclear fuel including the storage and transportation are investigated into interim storage, permanent disposal in geological repositories and overseas and domestic reprocessing. The ENVI code uses the GoldSim software to simulate the logistics of the associated activities. The result by the ENVI model not only produces the total cost to compare among the multiple options but also predict the sizes and timings of different facilities required. In order to decide the policy for spent nuclear management this purpose of this paper is to draw the optimum management plan to solve the nuclear spent fuel issue in the economical aspects. This paper is focused on the development of the ENVI's logic and calculations to simulate four options(No Reprocessing, Overseas Reprocessing, Domestic Reprocessing, and Overseas and Domestic Reprocessing) for managing the spent nuclear fuel in South Korea. The time history of the spent nuclear fuel produced from both the existing and future NPP's can be predicted, based on the Goldsim software made available very user friendly model. The simulation result will be used to suggest the strategic plans for the spent nuclear fuel management.
ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

International Nuclear Information System (INIS)

Chang, Sunyoung; Jeong, Yon Hong; Han, Jae-Jun; Lee, Aeri; Hwang, Yong-Soo

2015-01-01

The disposal facility of the spent nuclear fuel will be operated from 2051. This paper presents the ENVI code developed by GoldSim Software to simulate options for managing spent nuclear fuel (SNF) in South Korea. The ENVI is a simulator to allow decision-makers to assist to evaluate the performance for spent nuclear fuel management. The multiple options for managing the spent nuclear fuel including the storage and transportation are investigated into interim storage, permanent disposal in geological repositories and overseas and domestic reprocessing. The ENVI code uses the GoldSim software to simulate the logistics of the associated activities. The result by the ENVI model not only produces the total cost to compare among the multiple options but also predict the sizes and timings of different facilities required. In order to decide the policy for spent nuclear management this purpose of this paper is to draw the optimum management plan to solve the nuclear spent fuel issue in the economical aspects. This paper is focused on the development of the ENVI's logic and calculations to simulate four options(No Reprocessing, Overseas Reprocessing, Domestic Reprocessing, and Overseas and Domestic Reprocessing) for managing the spent nuclear fuel in South Korea. The time history of the spent nuclear fuel produced from both the existing and future NPP's can be predicted, based on the Goldsim software made available very user friendly model. The simulation result will be used to suggest the strategic plans for the spent nuclear fuel management
Historical overview of domestic spent fuel shipments: Update

International Nuclear Information System (INIS)

1991-07-01

This report presents available historic data on most commercial and research reactor spent fuel shipments in the United States from 1964 through 1989. Data include sources of the spent fuel shipped, types of shipping casks used, number of fuel assemblies shipped, and number of shipments made. This report also addresses the shipment of spent research reactor fuel. These shipments have not been documented as well as commercial power reactor spent fuel shipment activity. Available data indicate that the greatest number of research reactor fuel shipments occurred in 1986. The largest campaigns in 1986 were from the Brookhaven National Laboratory, Brooklyn, New York, to the Idaho Chemical Processing Plant (ICPP) and from the Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) in Tennessee and the Rockwell International Reactor in California to the Savannah River Plant near Aiken, South Carolina. For all years addressed in this report, DOE facilities in Idaho Falls and Savannah River were the major recipients of research reactor spent fuel. In 1989, 10 shipments were received at the Idaho facilities. These originated from universities in California, Michigan, and Missouri. 9 refs., 12 figs., 7 tabs
International safeguards concerns of Spent Fuel Disposal Program

International Nuclear Information System (INIS)

Pillay, K.K.S.

1988-01-01

The purpose of this paper is to stimulate discussions on the subjects of safeguarding large quantities of plutonium contained in spent fuels to be disposed of in geologic respositories. All the spent fuel disposal scenarios examined here pose a variety of safeguards problems, none of which are adequately addressed by the international safeguards community. The spent fuels from once-through fuel cycles in underground repositories would become an increasingly attractive target for diversion because of their plutonium content and decreasing radioactivity. Current design of the first geologic repository in the US will have the capacity to accommodate wastes equivalent to 70,000 Mt of uranium from commercial and defense fuel cycles. Of this, approximately 62,000 Mt uranium equivalent will be commerical spent fuel, containing over 500 Mt of plutonium. International safeguards commitments may require us to address the safeguards issues of disposing of such large quanities of plutonium in a geologic repository, which has the potential to become a plutonium mine in the future. This paper highlights several issues that should be addressed in the near term by US industries and the DOE before geologic repositories for spent fuels become a reality
Laser surveillance system for spent fuel

International Nuclear Information System (INIS)

Fiarman, S.; Zucker, M.S.; Bieber, A.M. Jr.

1980-01-01

A laser surveillance system installed at spent fuel storage pools (SFSP's) will provide the safeguard inspector with specific knowledge of spent fuel movement that cannot be obtained with current surveillance systems. The laser system will allow for the division of the pool's spent fuel inventory into two populations - those assemblies which have been moved and those which haven't - which is essential for maximizing the efficiency and effectiveness of the inspection effort. We have designed, constructed, and tested a full size laser system operating in air and have used an array of 6 zircaloy BWR tubes to simulate an assembly. The reflective signal from the zircaloy rods is a strong function of position of the assembly, but in all cases is easily discernable from the reference scan of the background with no assembly. A design for a SFSP laser surveillance system incorporating laser ranging is discussed. 10 figures
Data requirements and maintenance of records for spent fuel management: A review

International Nuclear Information System (INIS)

2006-11-01

Data collection and maintenance are an essential part of activities required in the lifetime management of spent fuel. Key data on spent fuel are required from the earliest phase of any project. To allow informed decisions for spent fuel management to be made, the data need to be maintained throughout the lifetime of spent fuel management including storage, transport, reprocessing or disposal. This publication is intended to provide a state-of-the-art review of spent fuel data management, including what data need to be gathered for the relevant activities in spent fuel management and how to maintain them by the responsible bodies at various stages of the nuclear fuel cycle. It provides some insights on a rational approach to spent fuel data management, considering the common requirements involved in spent fuel management for any Member State. In this regard, the information provided in these sections is mostly generic. After the introductory Section 1 and the Section 2 on data requirements for spent fuel management, Section 3 examines technical parameters that could specify spent fuel characteristics and associated conditions, followed by Section 4 on life cycle management of spent fuel data which includes the maintenance of records and other issues. Finally, some specific examples of the approaches already developed by a number of utilities and national organisations to characterise and track their spent-fuel data are presented in the Annex
Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR-III) 2009–2014

International Nuclear Information System (INIS)

2015-10-01

At the beginning of 2014, there were 437 nuclear power reactors in operation and 72 reactors under construction. To date, around 370 500 t (HM) (tonnes of heavy metal) of spent fuel have been discharged from reactors, and approximately 253 700 t (HM) are stored at various storage facilities. Although wet storage at reactor sites still dominates, the amount of spent fuel being transferred to dry storage technologies has increased significantly since 2005. For example, around 28% of the total fuel inventory in the United States of America is now in dry storage. Although the licensing for the construction of geological disposal facilities is under way in Finland, France and Sweden, the first facility is not expected to be available until 2025 and for most States with major nuclear programmes not for several decades afterwards. Spent fuel is currently accumulating at around 7000 t (HM) per year worldwide. The net result is that the duration of spent fuel storage has increased beyond what was originally foreseen. In order to demonstrate the safety of both spent fuel and the storage system, a good understanding of the processes that might cause deterioration is required. To address this, the IAEA continued the Coordinated Research Project (CRP) on Spent Fuel Performance Assessment and Research (SPAR-III) in 2009 to evaluate fuel and materials performance under wet and dry storage and to assess the impact of interim storage on associated spent fuel management activities (such as handling and transport). This has been achieved through: evaluating surveillance and monitoring programmes of spent fuel and storage facilities; collecting and exchanging relevant experience of spent fuel storage and the impact on associated spent fuel management activities; facilitating the transfer of knowledge by documenting the technical basis for spent fuel storage; creating synergy among research projects of the participating Member States; and developing the capability to assess the impact
A central spent fuel storage in Sweden

International Nuclear Information System (INIS)

Gustafsson, B.; Hagberth, R.

1978-01-01

A planned central spent fuel storage facility in Sweden is described. The nuclear power program and quantities of spent fuel generated in Sweden is discussed. A general description of the facility is given with emphasis on the lay-out of the buildings, transport casks and fuel handling. Finally a possible design of a Swedish transportation system is discussed. (author)
IEA-R1 reactor - Spent fuel management

International Nuclear Information System (INIS)

Mattos, J.R.L. De

1996-01-01

Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

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