Thermal and mechanical analyses of the spent nuclear fuel disposal canister and its barriers according to the design variable change

International Nuclear Information System (INIS)

Kwon, Young Joo

2006-03-01

This work constitutes a summary of research and development made for design and dimensioning of the spent nuclear fuel disposal canister. Since the spent nuclear fuel disposal emits high temperature heats and much radiation, its careful treatment is required. For that, a long term (usually 10,000 years) safe repository for the spent nuclear fuel disposal should be secured. Usually this repository is expected to locate at a depth of 500m underground. Many various analyses should be performed to secure the structural safety of the canister. For past years, these analyses have been performed to develop the canister model (so-called DKC-1 model). The diameter of the designed KDC-1 canister model is D=102m. However, there still remain some structural evaluations to make sure the structural safety of the designed KDC-1 canister mode. The one is the structural safety evaluation of the canister for the falling accident in the repository while handling the canister. There may happen two typical falling accidents in the repository. The one is the falling accident of the canister in the borehole while depositing the canister into the borehole. In these falling accidents the collision impact force between the canister and the surface of the ground or the bottom of the borehole may cause the structural damage onto the canister. However, the canister should be designed to withstand this impact force. Hence, the structural analysis of the canister for this impact force is required to guarantee the structural safety of the canister for this falling accident. Therefore in this report, the structural analyses of the KDC-1 canister model of the diameter of 102cm for two types of falling accidents are carried out for the impact forces while the canister collides onto the surface of the ground or the bottom of the borehole. The nonlinear structural analyses are performed for the canister to get the accurate analysis results assuming the materials composing canister parts as elasto

Impact Force Applied on the Spent Nuclear Fuel Disposal Canister that Accidentally Drops and Collides onto the Ground

International Nuclear Information System (INIS)

Kwon, Young Joo

2016-01-01

In this paper, a mathematical methodology was theoretically studied to obtain the impact force caused by the collision between rigid bodies. This theoretical methodology was applied to compute the impact force applied on the spent nuclear fuel disposal canister that accidentally drops and collides onto the ground. From this study, the impact force required to ensure a structurally safe canister design was theoretically formulated. The main content of the theoretical study concerns the rigid body kinematics and equation of motion during collision between two rigid bodies. On the basis of this study, a general impact theory to compute the impact force caused by the collision between two bodies was developed. This general impact theory was applied to theoretically formulate the approximate mathematical solution of the impact force that affects the spent nuclear fuel disposal canister that accidentally falls to the ground. Simultaneously, a numerical analysis was performed using the computer code to compute the numerical solution of the impact force, and the numerical result was compared with the approximate mathematical solution

Corrosion resistance of canisters for final disposal of spent nuclear fuel

International Nuclear Information System (INIS)

Mattsson, E.

1979-01-01

A group of Swedish scientists has evaluated from the corrosion point of view three alternative canister types for final disposal of waste from nuclear reactors in boreholes in rock 500 m below ground. Titanium canisters with a wall-thickness of 6 mm and 100 mm thick lead lining have been estimated to have a life of at least thousands of years, and probably tens of thousands of years. Copper canisters with 200-mm-thick walls would last for hundreds of thousands of years. The third type, α-alumina sintered under isostatic pressure, is a very promising canister material

Summary of Preliminary Criticality Analysis for Peach Bottom Fuel in the DOE Standardized Spent Nuclear Fuel Canister

International Nuclear Information System (INIS)

Henrikson, D.J.

1999-01-01

The Department of Energy's (DOE's) National Spent Nuclear Fuel Program is developing a standardized set of canisters for DOE spent nuclear fuel (SNF). These canisters will be used for DOE SNF handling, interim storage, transportation, and disposal in the national repository. Several fuels are being examined in conjunction with the DOE SNF canisters. This report summarizes the preliminary criticality safety analysis that addresses general fissile loading limits for Peach Bottom graphite fuel in the DOE SNF canister. The canister is considered both alone and inside the 5-HLW/DOE Long Spent Fuel Co-disposal Waste Package, and in intact and degraded conditions. Results are appropriate for a single DOE SNF canister. Specific facilities, equipment, canister internal structures, and scenarios for handling, storage, and transportation have not yet been defined and are not evaluated in this analysis. The analysis assumes that the DOE SNF canister is designed so that it maintains reasonable geometric integrity. Parameters important to the results are the canister outer diameter, inner diameter, and wall thickness. These parameters are assumed to have nominal dimensions of 45.7-cm (18.0-in.), 43.815-cm (17.25-in), and 0.953-cm (0.375-in.), respectively. Based on the analysis results, the recommended fissile loading for the DOE SNF canister is 13 Peach Bottom fuel elements if no internal steel is present, and 15 Peach Bottom fuel elements if credit is taken for internal steel

Efficiency analyses of the CANDU spent fuel repository using modified disposal canisters for a deep geological disposal system design

International Nuclear Information System (INIS)

Lee, J.Y.; Cho, D.K.; Lee, M.S.; Kook, D.H.; Choi, H.J.; Choi, J.W.; Wang, L.M.

2012-01-01

Highlights: ► A reference disposal concept for spent nuclear fuels in Korea has been reviewed. ► To enhance the disposal efficiency, alternative disposal concepts were developed. ► Thermal analyses for alternative disposal concepts were performed. ► From the result of the analyses, the disposal efficiency of the concepts was reviewed. ► The most effective concept was suggested. - Abstract: Deep geological disposal concept is considered to be the most preferable for isolating high-level radioactive waste (HLW), including nuclear spent fuels, from the biosphere in a safe manner. The purpose of deep geological disposal of HLW is to isolate radioactive waste and to inhibit its release of for a long time, so that its toxicity does not affect the human beings and the biosphere. One of the most important requirements of HLW repository design for a deep geological disposal system is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. In this study, a reference disposal concept for spent nuclear fuels in Korea has been reviewed, and based on this concept, efficient alternative concepts that consider modified CANDU spent fuels disposal canister, were developed. To meet the thermal requirement of the disposal system, the spacing of the disposal tunnels and that of the disposal pits for each alternative concept, were drawn following heat transfer analyses. From the result of the thermal analyses, the disposal efficiency of the alternative concepts was reviewed and the most effective concept suggested. The results of these analyses can be used for a deep geological repository design and detailed analyses, based on exact site characteristics data, will reduce the uncertainty of the results.

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

Disposal criticality analysis for aluminum-based DOE fuels

International Nuclear Information System (INIS)

Davis, J.W.; Gottlieb, P.

1997-11-01

This paper describes the disposal criticality analysis for canisters containing aluminum-based Department of Energy fuels from research reactors. Different canisters were designed for disposal of highly enriched uranium (HEU) and medium enriched uranium (MEU) fuel. In addition to the standard criticality concerns in storage and transportation, such as flooding, the disposal criticality analysis must consider the degradation of the fuel and components within the waste package. Massachusetts Institute of Technology (MIT) U-Al fuel with 93.5% enriched uranium and Oak Ridge Research Reactor (ORR) U-Si-Al fuel with 21% enriched uranium are representative of the HEU and MEU fuel inventories, respectively. Conceptual canister designs with 64 MIT assemblies (16/layer, 4 layers) or 40 ORR assemblies (10/layer, 4 layers) were developed for these fuel types. Borated stainless steel plates were incorporated into a stainless steel internal basket structure within a 439 mm OD, 15 mm thick XM-19 canister shell. The Codisposal waste package contains 5 HLW canisters (represented by 5 Defense Waste Processing Facility canisters from the Savannah River Site) with the fuel canister placed in the center. It is concluded that without the presence of a fairly insoluble neutron absorber, the long-term action of infiltrating water can lead to a small, but significant, probability of criticality for both the HEU and MEU fuels. The use of 1.5kg of Gd distributed throughout the MIT fuel and the use of carbon steels for the structural basket or 1.1 kg of Gd distributed in the ORR fuel will reduce the probability of criticality to virtually zero for both fuels

Mechanical failure of SKB spent fuel disposal canisters. Mathematical modelling and scoping calculations

International Nuclear Information System (INIS)

Takase, Hiroyasu; Benbow, S.; Grindrod, P.

1998-10-01

According to the current design of SKB, a copper overpack with a cast steel inner component will be used as the disposal canister for spent nuclear fuel. A recent study considered the case of a breach in the copper overpack, through which groundwater could enter the canister. It has pointed out that hydrogen gas generated by an anaerobic corrosion could cushion the system and reduce or eventually stop further infiltration of water into the breached canister, and thence the spent fuel. One potential pitfall in this previous study lies in the fact that it did not consider any processes which might violate the following assumptions which are essential for the gas 'cushioning': 1. Hydrogen gas accumulated in the annular gap in the canister forms a free gas phase which is stable indefinitely into future; 2. Elevated gas pressure in the canister prevents further supply of groundwater except for diffusion of vapour. In the current study we developed a set of mathematical models for the above problem and applied it to carry out an independent assessment of the long-term behaviour of the canister. A key aim in this study was to clarify whether there are any alternative processes which may affect the result obtained by the previous study by violating one of the assumptions listed above. For this purpose, a scenario development exercise was conducted. The result supported the concept described in the previous study. One exception is that possible intrusion of bentonite gel followed by its desaturation could leave paths both for the gas and water simultaneously without forming a gas cushion. This is summarised in the first part of the report. In the second part, development of mathematical models and their applications are described. The key results are: 1. The model describing behaviour of gas and pore water in the canister and the buffer material reproduced the main results of the previous study; 2. The model considering intrusion of the bentonite gel pointed out possibility

Storage, transportation and disposal system for used nuclear fuel assemblies

Science.gov (United States)

Scaglione, John M.; Wagner, John C.

2017-01-10

An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

Application of a cold spray technique to the fabrication of a copper canister for the geological disposal of CANDU spent fuels

Energy Technology Data Exchange (ETDEWEB)

Choi, Heui-Joo, E-mail: hjchoi@kaeri.re.k [Korea Atomic Energy Research Institute, Radioactive Waste Management Technology Development, 150 Dukjin-dong, Yuseong, Daejon, 305-353 (Korea, Republic of); Lee, Minsoo; Lee, Jong Youl [Korea Atomic Energy Research Institute, Radioactive Waste Management Technology Development, 150 Dukjin-dong, Yuseong, Daejon, 305-353 (Korea, Republic of)

2010-10-15

A new method was proposed for the manufacture of a copper-cast iron canister for the spent fuel disposal based on the cold spray coating technique. The thickness of a copper shell could be fabricated to be as thin as 10 mm with the new method. Around 6 tons of copper could be saved with a 10 mm thick canister compared with a 50 mm thick canister. The electrochemical properties of the cold sprayed copper layer and forged copper were measured through a polarization test. The two copper layers showed very similar electrochemical properties. The lifetime of a 10 mm copper canister was estimated with a mathematical model based on the mass transport of sulfide ions through the buffer. The results showed that the canister lifetime was more than 140,000 years under the Korean granite groundwater condition. The thermal analysis with a current pre-conceptual design of a CANDU spent fuel canister showed that the maximum temperature between the canister and the saturated buffer was below the thermal criteria, 100 {sup o}C. Finally, the mechanical stability of the copper canister was confirmed with a computer program, ABAQUS, under the rock movement scenario.

Interim transfer canister for consolidating nuclear fuel rods

International Nuclear Information System (INIS)

Formanek, F.J.

1987-01-01

This patent describes a canister for receiving and consolidating a group of uniformly spaced apart nuclear fuel rods, comprising: a rectangular, vertically oriented straight back panel; a pair of oppositely disposed side panels connected perpendicularly to the back panel, having a vertical straight upper portion and an inwardly tapered lower portion; a front panel opposite the back panel and connected to the side panels, having a straight vertical upper portion and inwardly tapered lower portion; whereby the back, side and front panels define a rectangular upper opening at the upper end of the canister and a generally rectangular lower opening at the other end, the lower opening having a cross-sectional area less than one-half that of the upper opening; parallel plate members spanning the canister from the front panel to the back panel, each plate spaced from the other the same uniform distance, the plates extending downwardly into the tapered portion of the canister while remaining spaced above the tapered sidewalls; first base means at the lower end of the canister, removably mounted and having an oblique orientation generally downward from the front panel to the back panel, for guiding the fuel rods to be inserted preferentially toward the lower portion of the back panel; and second base means removably mounted within the canister below first base means and oriented transversely to the longitudinal extent of the canister, for supporting the fuel rods when the first base means is removed from the canister

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

Canister Design for Deep Borehole Disposal of Nuclear Waste (CD-ROM)

National Research Council Canada - National Science Library

Hoag, Christopher I

2006-01-01

...: 1 CD-ROM; 4 3/4 in.; 28.7 MB. ABSTRACT: The objective of this thesis was to design a canister for the disposal of spent nuclear fuel and other high-level waste in deep borehole repositories using currently available and proven oil, gas...

Shaft shock absorber tests for a spent fuel canister

International Nuclear Information System (INIS)

Kukkola, T.; Toermaelae, V.P.

2005-06-01

The disposal canister for spent nuclear fuel will be transferred by a lift to the repository, which is 500 m deep in the bedrock. Model tests were carried out with the objective to estimate weather feasible shock absorber can be developed against the design accident case where the canister should survive a free fall to the lift shaft. If the velocity of the canister is not controlled by air drag or by any other deceleration means, the impact velocity may reach ultimate speed of 100m/s. The canister would retain its integrity in impact on water when the bottom pit of the lift well is filled with groundwater. However, the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20 m high filling to the bottom pit of the lift well by Light Expanded Clay Aggregate (LECA), gives fair impact absorption to protect the fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (orig.)

Shaft shock absorber tests for a spent fuel canister

International Nuclear Information System (INIS)

Kukkola, T.; Toermaelae, V.P.

2003-01-01

The holding canister for spent nuclear fuel will be transferred by a lift to the final disposal tunnels 500m deep in the bedrock. Model tests were carried out with an objective to estimate weather feasible shock absorbing properties can be met in a design accident case where the canister should survive a free fall due to e.g. sabotage. If the velocity of the canister is not controlled by air drag or any other deceleration means, the impact velocity may reach ultimate speed of 100m/s. The canister would retain its integrity when stricken by the surface penetration impact if the bottom pit of the lift well would be filled with groundwater. However the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20m high filling to the bottom pit of the lift well by ceramic gravel, trade mark LECA-sora, gives a fair impact absorption to protect the spent fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (orig.)

Alternative Concept to Enhance the Disposal Efficiency for CANDU Spent Fuel Disposal System

International Nuclear Information System (INIS)

Lee, Jong Youl; Cho, Dong Geun; Kook, Dong Hak; Lee, Min Soo; Choi, Heui Joo

2011-01-01

There are two types of nuclear reactors in Korea and they are PWR type and CANDU type. The safe management of the spent fuels from these reactors is very important factor to maintain the sustainable energy supply with nuclear power plant. In Korea, a reference disposal system for the spent fuels has been developed through a study on the direct disposal of the PWR and CANDU spent fuel. Recently, the research on the demonstration and the efficiency analyses of the disposal system has been performed to make the disposal system safer and more economic. PWR spent fuels which include a lot of reusable material can be considered being recycled and a study on the disposal of HLW from this recycling process is being performed. CANDU spent fuels are considered being disposed of directly in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System (KRS) which was to dispose of both PWR type and CANDU type, the more effective CANDU spent fuel disposal systems were developed. To do this, the disposal canister for CANDU spent fuels was modified to hold the storage basket for 60 bundles which is used in nuclear power plant. With these modified disposal canister concepts, the disposal concepts to meet the thermal requirement that the temperature of the buffer materials should not be over 100 .deg. C were developed. These disposal concepts were reviewed and analyzed in terms of disposal effective factors which were thermal effectiveness, U-density, disposal area, excavation volume, material volume etc. and the most effective concept was proposed. The results of this study will be used in the development of various wastes disposal system together with the HLW wastes from the PWR spent fuel recycling process.

Cost estimate of Olkiluoto disposal facility for spent nuclear fuel

International Nuclear Information System (INIS)

Kukkola, T.; Saanio, T.

2005-03-01

The cost estimate covers the underground rock characterisation facility ONKALO, the investment and the operating costs of the above and underground facilities, the decommissioning of the encapsulation plant and the closure costs of the repository. The above ground facility is a once-investment; a re-investment takes place after 37 years operation. The repository is extended stepwise thus also the investment take place in stages. Annual operating costs are calculated with different operating efficiencies. The total investment costs of the disposal facility are estimated to be 503 M euro (Million Euros), the total operating costs are 1,923 M euro and the decommissioning and the closure costs are 116 M euro totaling 2,542 M euro. The investment costs of the above ground facility are 142 M euro, the operating costs are 1,678 M euro. The repository investment costs are 360 M euro and the operating costs are 245 M euro. The decommissioning costs are 7 M euro and the closure costs are 109 M euro. The costs are calculated by using the price level of December 2003. The cost estimate is based on a plan, where the spent fuel is encapsulated and the disposal canisters are disposed into the bedrock at a depth of about 420 meters in one storey. In the encapsulation process, the fuel assemblies are closed into composite canisters, in which the inner part of the canister is made of nodular cast iron and the outer wall of copper having a thickness of 50 mm. The inner canister is closed gas-tight by a bolted steel lid, and the electron beam welding method is used to close the outer copper lid. The encapsulation plant is independent and located above the deep repository spaces. The disposal canisters are transported to the repository by the lift. The disposal tunnels are constructed and closed in stages according the disposal canisters disposal. The operating time of the Loviisa nuclear power plant units is assumed to be 50 years and the operating time of the Olkiluoto nuclear power

Performance Assessment and Sensitivity Analyses of Disposal of Plutonium as Can-in-Canister Ceramic

International Nuclear Information System (INIS)

Rainer Senger

2001-01-01

The purpose of this analysis is to examine whether there is a justification for using high-level waste (HLW) as a surrogate for plutonium disposal in can-in-canister ceramic in the total-system performance assessment (TSPA) model for the Site Recommendation (SR). In the TSPA-SR model, the immobilized plutonium waste form is not explicitly represented, but is implicitly represented as an equal number of canisters of HLW. There are about 50 metric tons of plutonium in the U. S. Department of Energy inventory of surplus fissile material that could be disposed. Approximately 17 tons of this material contain significant quantities of impurities and are considered unsuitable for mixed-oxide (MOX) reactor fuel. This material has been designated for direct disposal by immobilization in a ceramic waste form and encapsulating this waste form in high-level waste (HLW). The remaining plutonium is suitable for incorporation into MOX fuel assemblies for commercial reactors (Shaw 1999, Section 2). In this analysis, two cases of immobilized plutonium disposal are analyzed, the 17-ton case and the 13-ton case (Shaw et al. 2001, Section 2.2). The MOX spent-fuel disposal is not analyzed in this report. In the TSPA-VA (CRWMS M and O 1998a, Appendix B, Section B-4), the calculated dose release from immobilized plutonium waste form (can-in-canister ceramic) did not exceed that from an equivalent amount of HLW glass. This indicates that the HLW could be used as a surrogate for the plutonium can-in-canister ceramic. Representation of can-in-canister ceramic as a surrogate is necessary to reduce the number of waste forms in the TSPA model. This reduction reduces the complexity and running time of the TSPA model and makes the analyses tractable. This document was developed under a Technical Work Plan (CRWMS M and O 2000a), and is compliant with that plan. The application of the Quality Assurance (QA) program to the development of that plan (CRWMS M and O 2000a) and of this Analysis is

Assessment of a spent fuel disposal canister. Assessment studies for a copper canister with cast steel inner component

International Nuclear Information System (INIS)

Bond, A.E.; Hoch, A.R.; Jones, G.D.; Tomczyk, A.J.; Wiggin, R.M.; Worraker, W.J.

1997-05-01

The proposed design for a final repository for spent fuel and other long-lived residues in Sweden, is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in vertical storage holes drilled in a series of caverns excavated from the granite bedrock at a depth of about 500 m. Each canister will be surrounded by compacted bentonite clay. In this report, a simple model of the behaviour of the canister subsequent to a first breach in its copper overpack is developed. This model is used to predict: -the ingress of water to the canister (as a function of the size and the shape of the initial defect, the buffer conductivity, the corrosion rate and the pressure inside the canister); -the build-up of corrosion products in the canister (as a function of the available water in the canister, the corrosion rate and the properties of the corrosion products); -the effect of corrosion on the structural integrity of the canister. A number of different scenarios for the location of the breach in the copper overpack are considered

A review of materials and corrosion issues regarding canisters for disposal of spent fuel and high-level waste in Opalinus clay

International Nuclear Information System (INIS)

Landolt, D.; Davenport, A.; Payer, J.; Shoesmith, D.

2009-01-01

The project 'Entsorgungsnachweis' presented by NAGRA to the Swiss Federal Government in December 2002 assessed the feasibility of disposal of spent fuel (SF), vitrified high level waste (HLW) from reprocessing and long-lived intermediate level waste in an Opalinus Clay repository site in Northern Switzerland. NAGRA proposed the use of carbon steel canisters for disposal of SF/HLW and it also put forward an alternative concept of copper canisters with cast iron insert. In its reply the Federal Government acknowledged that NAGRA had successfully demonstrated the technical feasibility of disposal of SF/HLW. However, some of its experts raised a number of questions related to the choice of steel as canister material. Among others, it was questioned whether hydrogen formed by corrosion of steel in contact with saturated bentonite might adversely affect the barrier function of the Opalinus clay. It was also recommended that alternative canister materials and/or design concepts should be evaluated. To deal with these concerns NAGRA convened an international group of experts, the Canister Materials Review Board (CMRB), who were to review the existing information on canister materials that could be suitable for the proposed repository environment. Based on present knowledge of materials science, the CMRB was to recommend to NAGRA the most suitable material(s) for meeting the performance requirements for SF/HLW canisters. Specifically, the CMRB was to consider corrosion, including hydrogen generation, and stress-assisted failure processes that could affect the integrity and projected life time of SF/HLW canisters or impede the functioning of geological barriers while keeping in mind the overall feasibility of manufacturing, sealing and inspecting the canisters. The CMRB was further asked to identify the needs and provide advice for further studies by NAGRA on the long term performance and safety of SF/HLW canisters in the Swiss repository concept. For the assessment of the

Safety analysis of disposal of spent nuclear fuel

International Nuclear Information System (INIS)

Vieno, T.

1994-04-01

The spent fuel from the Olkiluoto NPP (TVO I and II) is planned to be disposed of in a repository to be constructed at a depth of about 500 meters in the crystalline bedrock. The thesis is dealing with the safety analysis of the disposal. The main topics presented in the thesis are: (1) The amount of radioactive properties of the spent fuel, (2) The canister design and the planned disposal concept, (3) The results of the preliminary site investigations, (4) Discussion of the multi-barrier principle, (5) The general principles and methodology of the TVO-92 safety analysis, (6) Groundwater flow analysis, (7) Durability and behaviour of the canister, (8) Biosphere analysis and reference scenario, and (9) The sensitivity and uncertainty analyses. (246 refs., 75 figs., 44 tabs.)

Radiological consequences of accidents during disposal of spent nuclear fuel in a deep borehole

Energy Technology Data Exchange (ETDEWEB)

Grundfelt, Bertil [Kemakta Konsult AB, Stockholm (Sweden)

2013-07-15

In this report, an analysis of the radiological consequences of potential accidents during disposal of spent nuclear fuel in deep boreholes is presented. The results presented should be seen as coarse estimates of possible radiological consequences of a canister being stuck in a borehole during disposal rather than being the results of a full safety analysis. In the concept for deep borehole disposal of spent nuclear fuel developed by Sandia National Laboratories, the fuel is assumed to be encapsulated in mild steel canisters and stacked between 3 and 5 km depth in boreholes that are cased with perforated mild steel casing tubes. The canisters are joined together by couplings to form strings of 40 canisters and lowered into the borehole. When a canister string has been emplaced in the borehole, a bridge plug is installed above the string and a 10 metres long concrete plug is cast on top of the bridge plug creating a floor for the disposal of the next sting. In total 10 canister strings, in all 400 canisters, are assumed to be disposed of at between 3 and 5 kilometres depth in one borehole. An analysis of potential accidents during the disposal operations shows that the potentially worst accident would be that a canister string is stuck above the disposal zone of a borehole and cannot be retrieved. In such a case, the borehole may have to be sealed in the best possible way and abandoned. The consequences of this could be that one or more leaking canisters are stuck in a borehole section with mobile groundwater. In the case of a leaking canister being stuck in a borehole section with mobile groundwater, the potential radiological consequences are likely to be dominated by the release of the so-called Instant Release Fraction (IRF) of the radionuclide inventory, i.e. the fraction of the radionuclides that as a consequence of the in-core conditions are present in the annulus between the fuel pellets and the cladding or on the grain boundaries of the UO{sub 2} matrix

Radiological consequences of accidents during disposal of spent nuclear fuel in a deep borehole

International Nuclear Information System (INIS)

Grundfelt, Bertil

2013-07-01

In this report, an analysis of the radiological consequences of potential accidents during disposal of spent nuclear fuel in deep boreholes is presented. The results presented should be seen as coarse estimates of possible radiological consequences of a canister being stuck in a borehole during disposal rather than being the results of a full safety analysis. In the concept for deep borehole disposal of spent nuclear fuel developed by Sandia National Laboratories, the fuel is assumed to be encapsulated in mild steel canisters and stacked between 3 and 5 km depth in boreholes that are cased with perforated mild steel casing tubes. The canisters are joined together by couplings to form strings of 40 canisters and lowered into the borehole. When a canister string has been emplaced in the borehole, a bridge plug is installed above the string and a 10 metres long concrete plug is cast on top of the bridge plug creating a floor for the disposal of the next sting. In total 10 canister strings, in all 400 canisters, are assumed to be disposed of at between 3 and 5 kilometres depth in one borehole. An analysis of potential accidents during the disposal operations shows that the potentially worst accident would be that a canister string is stuck above the disposal zone of a borehole and cannot be retrieved. In such a case, the borehole may have to be sealed in the best possible way and abandoned. The consequences of this could be that one or more leaking canisters are stuck in a borehole section with mobile groundwater. In the case of a leaking canister being stuck in a borehole section with mobile groundwater, the potential radiological consequences are likely to be dominated by the release of the so-called Instant Release Fraction (IRF) of the radionuclide inventory, i.e. the fraction of the radionuclides that as a consequence of the in-core conditions are present in the annulus between the fuel pellets and the cladding or on the grain boundaries of the UO 2 matrix. The

Criticality safety considerations in the geologic disposal of spent nuclear fuel assemblies

International Nuclear Information System (INIS)

Gore, B.F.; McNair, G.W.; Heaberlin, S.W.

1980-05-01

Features of geologic disposal which hamper the demonstration that criticality cannot occur therein include possible changes of shape and form, intrusion of water as a neutron moderator, and selective leaching of spent fuel constituents. If the criticality safety of spent fuel disposal depends on burnup, independent measurements verifying the burnup should be performed prior to disposal. The status of nondestructive analysis method which might provide such verification is discussed. Calculations were performed to assess the potential for increasing the allowed size of a spent fuel disposal canister if potential water intrusion were limited by close-packing the enclosed rods. Several factors were identified which severely limited the potential of this application. The theoretical limit of hexagonal close-packing cannot be achieved due to fuel rod bowing. It is concluded that disposal canisters should be sized on the basis of assumed optimum moderation. Several topics for additional research were identified during this limited study

Modeling of thermal evolution of near field area around single pit mode nuclear waste canister disposal in soft rocks

International Nuclear Information System (INIS)

Bajpai, R.K.; Verma, A.K.; Maheshwar, Sachin

2016-01-01

Soft rocks like argillites/shales are under consideration worldwide as host rock for geological disposal of vitrified as well as spent fuel nuclear waste. The near field around disposed waste canister at 400-500m depth witnesses a complex heat field evolution due to varying thermal characteristics of rocks, coupling with hydraulic processes and varying intensity of heat flux from the canister. Smooth heat dissipation across the rock is desirable to avoid buildup of temperature beyond design limit (100 °C) and resultant micro fracturing due to thermal stresses in the rocks and intervening buffer clay layers. This also causes enhancement of hydraulic conductivity of the rocks, radionuclide transport and greater groundwater ingress towards the canister. Hence heat evolution modeling constitutes an important part of safety assessment of geological disposal facilities

Production methods and costs of oxygen free copper canisters for nuclear waste disposal

International Nuclear Information System (INIS)

Aalto, H.; Rajainmaeki, H.; Laakso, L.

1996-10-01

The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for disposal of spent nuclear fuel from reactors of Teollisuuden Voima Oy (TVO) and Imatran Voima Oy (IVO) are discussed. The canister design is based on the Posiva's concept where solid insert structure is surrounded by the copper mantle. During recent years Outokumpu Copper Products and Posiva have continued their work on development of the copper canisters. Outokumpu Copper Products has also increased capability to manufacture these canisters. In the study the most potential manufacturing methods and their costs are discussed. The cost estimates are based on the assumption that Outokumpu will supply complete copper mantles. At the moment there are at least two commercially available production methods for copper cylinder manufacturing. These routes are based on either hot extrusion of the copper tube or hot rolling, bending and EB-welding of the tube. Trial fabrications has been carried out with both methods for the full size canisters. These trials of the canisters has shown that both the forming from rolled plate and the extrusion are possible methods for fabricating copper canisters on a full scale. (orig.) (26 refs.)

Deep Borehole Disposal Concept: Development of Universal Canister Concept of Operations

Energy Technology Data Exchange (ETDEWEB)

Rigali, Mark J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Applied Systems Analysis and Research; Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Applied Systems Analysis and Research

2016-08-01

This report documents key elements of the conceptual design for deep borehole disposal of radioactive waste to support the development of a universal canister concept of operations. A universal canister is a canister that is designed to be able to store, transport, and dispose of radioactive waste without the canister having to be reopened to treat or repackage the waste. This report focuses on the conceptual design for disposal of radioactive waste contained in a universal canister in a deep borehole. The general deep borehole disposal concept consists of drilling a borehole into crystalline basement rock to a depth of about 5 km, emplacing WPs in the lower 2 km of the borehole, and sealing and plugging the upper 3 km. Research and development programs for deep borehole disposal have been ongoing for several years in the United States and the United Kingdom; these studies have shown that deep borehole disposal of radioactive waste could be safe, cost effective, and technically feasible. The design concepts described in this report are workable solutions based on expert judgment, and are intended to guide follow-on design activities. Both preclosure and postclosure safety were considered in the development of the reference design concept. The requirements and assumptions that form the basis for the deep borehole disposal concept include WP performance requirements, radiological protection requirements, surface handling and transport requirements, and emplacement requirements. The key features of the reference disposal concept include borehole drilling and construction concepts, WP designs, and waste handling and emplacement concepts. These features are supported by engineering analyses.

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)

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)

Remote controlled mover for disposal canister transfer

International Nuclear Information System (INIS)

Suikki, M.

2013-10-01

This working report is an update for an earlier automatic guided vehicle design (Pietikaeinen 2003). The short horizontal transfers of disposal canisters manufactured in the encapsulation process are conducted with remote controlled movers both in the encapsulation plant and in the underground areas at the canister loading station of the disposal facility. The canister mover is a remote controlled transfer vehicle mobile on wheels. The handling of canisters is conducted with the assistance of transport platforms (pallets). The very small automatic guided vehicle of the earlier design was replaced with a commercial type mover. The most important reasons for this being the increased loadbearing requirement and the simpler, proven technology of the vehicle. The larger size of the vehicle induced changes to the plant layouts and in the principles for dealing with fault conditions. The selected mover is a vehicle, which is normally operated from alongside. In this application, the vehicle steering technology must be remote controlled. In addition, the area utilization must be as efficient as possible. This is why the vehicle was downsized in its outer dimensions and supplemented with certain auxiliary equipment and structures. This enables both remote controlled operation and improves the vehicle in terms of its failure tolerance. Operation of the vehicle was subjected to a risk analysis (PFMEA) and to a separate additional calculation conserning possible canister toppling risks. The total cost estimate, without value added tax for manufacturing the system amounts to 730 000 euros. (orig.)

Remote controlled mover for disposal canister transfer

Energy Technology Data Exchange (ETDEWEB)

Suikki, M. [Optimik Oy, Turku (Finland)

2013-10-15

This working report is an update for an earlier automatic guided vehicle design (Pietikaeinen 2003). The short horizontal transfers of disposal canisters manufactured in the encapsulation process are conducted with remote controlled movers both in the encapsulation plant and in the underground areas at the canister loading station of the disposal facility. The canister mover is a remote controlled transfer vehicle mobile on wheels. The handling of canisters is conducted with the assistance of transport platforms (pallets). The very small automatic guided vehicle of the earlier design was replaced with a commercial type mover. The most important reasons for this being the increased loadbearing requirement and the simpler, proven technology of the vehicle. The larger size of the vehicle induced changes to the plant layouts and in the principles for dealing with fault conditions. The selected mover is a vehicle, which is normally operated from alongside. In this application, the vehicle steering technology must be remote controlled. In addition, the area utilization must be as efficient as possible. This is why the vehicle was downsized in its outer dimensions and supplemented with certain auxiliary equipment and structures. This enables both remote controlled operation and improves the vehicle in terms of its failure tolerance. Operation of the vehicle was subjected to a risk analysis (PFMEA) and to a separate additional calculation conserning possible canister toppling risks. The total cost estimate, without value added tax for manufacturing the system amounts to 730 000 euros. (orig.)

Drop of canistered spent fuel segments into a deep borehole and subsequent aerosol release

International Nuclear Information System (INIS)

Bantle, S.; Herbe, H.; Miu, J.

1991-09-01

The source term of the released aerosols is estimated. First, the number of failing canisters is calculated for the case of an axial symmetric canister (POLLUX) pile, and then, for the case of a 'zig-zag' pile, as found in reality. The weight-specific energy acting on the fuel - a measure for the degree of fuel fractioning - is determined from the acceleration acting on the pin segments. In the borehole prevails a steady-state flow pattern which is stimulated by the heat of the disposed waste canister, and is also influenced by the ventilation of the drift above the borehole. Based on this stationary flow pattern flow velocities are calculated by means of fluid mechanical methods. Further investigations deal with the unsteady case which occurs during and immediately after the canister drop as well as with the wake behind the canister. The most relevant result is that under the considered boundary conditions no release form the borehole into the repository is to be expected. (orig./HP) [de

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

Canister arrangement for storing radioactive waste

Science.gov (United States)

Lorenzo, D.K.; Van Cleve, J.E. Jr.

1980-04-23

The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

Performance Specification Shippinpark Pressurized Water Reactor Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shippingport Spent Fuel Canisters

International Nuclear Information System (INIS)

JOHNSON, D.M.

2000-01-01

This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders

Rates and mechanisms of radioactive release and retention inside a waste disposal canister - in Can Processes

Energy Technology Data Exchange (ETDEWEB)

Oversby, V.M. (ed.) [and others

2003-10-01

Sweden and Finland are planning to dispose of spent nuclear fuel in a deep underground repository constructed in granitic rock. Each country is investigating candidate sites and developing the scientific and technical basis for assessing the safety of an eventual repository. An essential part of the safety assessment involves understanding the behaviour of the spent fuel after it is placed in the geologic environment. The fuel will be sealed inside a copper canister that contains a cast iron insert. The copper functions as a corrosion resistant barrier, while the cast iron insert fills much of the internal void space, adding strength to the canister and reducing the space available for water to accumulate inside the canister after the corrosion barrier is breached. The canisters will be surrounded by compressed bentonite, which will limit the access of water and dissolved species to the canister. Oxygen that is initially present when the disposal environment is sealed will be rapidly consumed by pyrite in the bentonite, bacterial species in the rock, and reduced inorganic materials in the rock. The copper canister will prevent access of water to the iron until it is corroded through, a process that is expected to take millions of years. After water contacts the iron, anaerobic corrosion of the insert will generate hydrogen gas and introduce Fe(II) ions into the water. The long-term environment for the fuel, therefore, is a highly reducing environment. The only possible source of oxidising agents is radiolysis of the water by radiation from the fuel. In the long-term, the radioactivity in the fuel is due to isotopes that decay by alpha decay; most of the activity from beta and gamma radiation will have decayed away. Spent fuel that is available for testing contains high levels of beta and gamma activity. Even when testing is done in the presence of hydrogen or actively corroding iron, the radiolysis due to beta and gamma radiation can introduce oxidising agents into

Volumes, Masses, and Surface Areas for Shippingport LWBR Spent Nuclear Fuel in a DOE SNF Canister

International Nuclear Information System (INIS)

J.W. Davis

1999-01-01

The purpose of this calculation is to estimate volumes, masses, and surface areas associated with (a) an empty Department of Energy (DOE) 18-inch diameter, 15-ft long spent nuclear fuel (SNF) canister, (b) an empty DOE 24-inch diameter, 15-ft long SNF canister, (c) Shippingport Light Water Breeder Reactor (LWBR) SNF, and (d) the internal basket structure for the 18-in. canister that has been designed specifically to accommodate Seed fuel from the Shippingport LWBR. Estimates of volumes, masses, and surface areas are needed as input to structural, thermal, geochemical, nuclear criticality, and radiation shielding calculations to ensure the viability of the proposed disposal configuration

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).

Canisters and nonfuel components at commercial nuclear reactors

International Nuclear Information System (INIS)

Gibbard, K.; Disbrow, J.

1994-01-01

This paper discusses detailed data on canisters and nonfuel components (NFC) at US commercial nuclear power reactors. A wide variety of NFC have been reported on the Form RW-859, open-quotes Nuclear Fuel Dataclose quotes survey. They may have been integral with an assembly, noncanistered in baskets, destined for disposal as low-level radioactive waste, or stored in canisters. Similarly, data on the family of canistered spent nuclear fuel (SNF) in storage pools was compiled. Approximately 85 percent of the 40,194 pieces of nonfuel assembly (NFA) hardware reported were integral with an assembly. This represents data submitted by 95 of the 107 reactors in 10 generic assembly classes. In addition, a total of 286 canisters have been reported as being in storage pools as of December 31, 1992. However, an additional 264 open baskets were also reported to contain miscellaneous SNF and nonfuel materials, garbage and debris. All of these 286 canisters meet the dimensional envelope requirements specified for disposal for open-quotes standard fuelclose quotes under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR 961); most of the baskets do not

Effect of canister size on costs of disposal of SRP high-level wastes

International Nuclear Information System (INIS)

McDonell, W.R.

1982-01-01

The current plan for managing the high-level nuclear wastes at the Savannah River Plant (SRP) calls for processing them into solid forms contained in stainless steel canisters for eventual disposal in a federal geologic repository. A new SRP facility called the Defense Waste Processing Facility (DWPF) is being designed for the onsite waste processing operations. Preliminary evaluations indicate that costs of the overall disposal operation will depend significantly on the size of the canisters, which determines the number of waste forms to be processed. The objective of this study was to evaluate the effects of canister size on costs of DWPF process operations, including canister procurement, waste solidification, and interim storage, on offsite transport, and on repository costs of disposal, including provision of suitable waste packages

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).

Final disposal of spent fuel in the Finnish bedrock

International Nuclear Information System (INIS)

1992-12-01

Teollisuuden Voima Oy (TVO) is preparing for the final disposal of spent nuclear fuel from the Olkiluoto nuclear power plant (TVO-I and TVO-II reactors). According to present estimates, a total of 1840 tU of spent fuel will be accumulated during the 40-year lifetime of the power plant. An interim storage facility for spent fuel (TVO-KPA Store) has operated at Olkiluoto since 1987. The spent fuel will be held in storage for several decades before it is shipped to the repository site. Both train and road transportation are possible. The spent fuel will be encapsulated in composite copper and steel canisters (ACP Canister) in a facility that will be build above the ground on the site where the repository is located. The repository will be constructed at the depth of several hundreds of meters in the bedrock. In 1987 five areas were selected for preliminary site investigations. The safety analysis (TVO-92) that was carried out shows that the proposed safety criteria would be met at each of the candidate sites. In future expected conditions there would never be significant releases of radioactive substances to the biosphere. The site investigations will be continued in the period 1993 to 2000. In parallel, a R and D programme will be devoted to the safety and technology of final disposal. The site for final disposal will be selected in the year 2000 with the aim of having the capability to start the disposal operations in 2020

Development of ultrasonic immersion inspection technique for spent fuel canisters

International Nuclear Information System (INIS)

Schankula, J.J.

1982-07-01

This report summarizes ultrasonic nondestructive testing development for metal matrix supported spent fuel disposal canisters. The work has concentated in two areas: inspection for lack of bond at the shell/matrix interface and inspection for voids in the matrix. The capabilities and limitations of these techniques have been fully established. Unbonded areas as small as 4 mm in diameter and voids 6 mm in diameter, 25 mm deep in the matrix, can readily be detected

The design analysis of ACP-canister for nuclear waste disposal

International Nuclear Information System (INIS)

Raiko, H.

1992-05-01

The design basis, dimensioning and some manufacturing aspects of the Advanced Cold Process Canister (ACPC) for the nuclear waste disposal is summarized in the report. The strength of the canister has been evaluated in normal design load condition and in extreme high hydrostatic pressure load condition possibly caused by ice age (orig.)

Development of geological disposal system for spent fuels and high-level radioactive wastes in Korea

International Nuclear Information System (INIS)

Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won

2013-01-01

Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel) for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

Feasibility study for a DOE research and production fuel multipurpose canister

International Nuclear Information System (INIS)

Lopez, D.A.; Abbott, D.G.

1994-02-01

This is a report of the feasibility of multipurpose canisters for transporting, storing, and sing of Department of Energy research and production spent nuclear fuel. Six representative Department of Energy fuel assemblies were selected, and preconceptual canister designs were developed to accommodate these assemblies. The study considered physical interface, structural adequacy, criticality safety, shielding capability, thermal performance of the canisters, and fuel storage site infrastructure. The external envelope of the canisters was designed to fit within the overpack casks for commercial canisters being developed for the Department of Energy Office of Civilian Radioactive Waste Management. The budgetary cost of canisters to handle all fuel considered is estimated at $170.8M. One large conceptual boiling water reactor canister design, developed for the Office of Civilian Radioactive Waste Management, and two new canister designs can accommodate at least 85% of the volume of the Department of Energy fuel considered. Canister use minimizes public radiation exposure and is cost effective compared with bare fuel handling. Results suggest the need for additional study of issues affecting canister use and for conceptual design development of the three canisters

Canister Cleaning System Final Design Report - Project A.2.A

International Nuclear Information System (INIS)

FARWICK, C.C.

2000-01-01

Approximately 2,300 metric tons Spent Nuclear Fuel (SNF) are currently stored within two water filled pools, the 105 K East (KE) fuel storage basin and the 105 K West (KW) fuel storage basin, at the U.S. Department of Energy, Richland Operations Office (RL). The SNF Project is responsible for operation of the K Basins and for the materials within them. A subproject to the SNF Project is the Debris Removal Subproject, which is responsible for removal of empty canisters and lids from the basins. The Canister Cleaning System (CCS) is part of the Debris Removal Project. The CCS will be installed in the KW Basin and operated during the fuel removal activity. The KW Basin has approximately 3600 canisters that require removal from the basin. The CCS is being designed to ''clean'' empty fuel canisters and lids and package them for disposal to the Environmental Restoration Disposal Facility complex. The system will interface with the KW Basin and be located in the Dummy Elevator Pit

Choices of canisters and elements for the first fuel and canister sludge shipment from K East Basin

International Nuclear Information System (INIS)

Makenas, B.J.

1996-01-01

The K East Basin contains open-top canisters with up to fourteen N Reactor fuel assemblies distributed between the two barrels of each canister. Each fuel assembly generally consists of inner and outer concentric elements fabricated from uranium metal with zirconium alloy cladding. The canisters also contain varying amounts of accumulated sludge. Retrieval of sample fuel elements and associated sludge for examination is scheduled to occur in the near future. The purpose of this document is to specify particular canisters and elements of interest as candidate sources of fuel and sludge to be shipped to laboratories

Desludging of N Reactor fuel canisters: Analysis, Test, and data requirements

International Nuclear Information System (INIS)

Johnson, A.B. Jr.

1996-01-01

The N Reactor fuel is currently stored in canisters in the K East (KE) and K West (KW) Basins. In KE, the canisters have open tops; in KW, the cans have sealed lids, but are vented to release gases. Corrosion products have formed on exposed uranium metal fuel, on carbon steel basin component surfaces, and on aluminum alloy canister surfaces. Much of the corrosion product is retained on the corroding surfaces; however, large inventories of particulates have been released. Some of the corrosion product particulates form sludge on the basin floors; some particulates are retained within the canisters. The floor sludge inventories are much greater in the KE Basin than in the KW Basin because KE Basin operated longer and its water chemistry was less controlled. Another important factor is the absence of lids on the KE canisters, allowing uranium corrosion products to escape and water-borne species, principally iron oxides, to settle in the canisters. The inventories of corrosion products, including those released as particulates inside the canisters, are only beginning to be characterized for the closed canisters in KW Basin. The dominant species in the KE floor sludge are oxides of aluminum, iron, and uranium. A large fraction of the aluminum and uranium floor sludge particulates may have been released during a major fuel segregation campaign in the 1980s, when fuel was emptied from 4990 canisters. Handling and jarring of the fuel and aluminum canisters seems likely to have released particulates from the heavily corroded surfaces. Four candidate methods are discussed for dealing with canister sludge emerged in the N Reactor fuel path forward: place fuel in multi-canister overpacks (MCOs) without desludging; drill holes in canisters and drain; drill holes in canisters and flush with water; and remove sludge and repackage the fuel

Structural Integrity Evaluation for Damaged Fuel Canister of a Research Reactor

International Nuclear Information System (INIS)

Oh, Jinho; Kwak, Jinsung; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo

2016-01-01

The purpose of this document is to confirm the structural integrity of damaged fuel canister through the numerical simulation. The analysis results of canister including damaged fuel are evaluated with design limits of the ASME Sec. III NF Codes and Standards. The main function of canister is to store and protect the damaged fuel assembly generated from the operation of the research reactor. The canister is classified into safety class NNS (Non-nuclear Safety) and seismic category II. The shape of the canister is designed into commercialized circular tube due to economic benefit and easy manufacturing. The damaged fuel assembly is loaded in a dedicated canister by using special tool and supported by lower block in the canister. Then it is move into the damaged fuel storage rack under safeguards arrangements. The canister is securely supported at guide plate and base plate of rack. The structural integrity evaluation for the canister is performed by using response spectrum analysis. The analysis results show that the stress intensity of the canister under the seismic loads is within the ASME Code limits. Thus, the validity of the present design of the canister has been demonstrated

Structural Integrity Evaluation for Damaged Fuel Canister of a Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Oh, Jinho; Kwak, Jinsung; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The purpose of this document is to confirm the structural integrity of damaged fuel canister through the numerical simulation. The analysis results of canister including damaged fuel are evaluated with design limits of the ASME Sec. III NF Codes and Standards. The main function of canister is to store and protect the damaged fuel assembly generated from the operation of the research reactor. The canister is classified into safety class NNS (Non-nuclear Safety) and seismic category II. The shape of the canister is designed into commercialized circular tube due to economic benefit and easy manufacturing. The damaged fuel assembly is loaded in a dedicated canister by using special tool and supported by lower block in the canister. Then it is move into the damaged fuel storage rack under safeguards arrangements. The canister is securely supported at guide plate and base plate of rack. The structural integrity evaluation for the canister is performed by using response spectrum analysis. The analysis results show that the stress intensity of the canister under the seismic loads is within the ASME Code limits. Thus, the validity of the present design of the canister has been demonstrated.

DEVELOPMENT OF GEOLOGICAL DISPOSAL SYSTEMS FOR SPENT FUELS AND HIGH-LEVEL RADIOACTIVE WASTES IN KOREA

Directory of Open Access Journals (Sweden)

HEUI-JOO CHOI

2013-02-01

Full Text Available Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

Human health considerations in the assessment of Canadian concept for the disposal of nuclear fuel wastes

International Nuclear Information System (INIS)

Baweja, A.S.; Tracy, B.L.; Ahier, B.; Bartlett, S.

1996-01-01

In 1978, AECL was mandated by the government of Ontario and the federal government to find a permanent disposal solution for spent nuclear fuels. Canada opted for disposal in plutonic rocks of the Canadian shield. The Canadian concept calls for disposal in crystalline rocks at a depth of 500 to 1000 m below the surface. The spent fuel would be contained in a canister, the canister would be emplaced in a vault containing clay-based buffer materials, and the cavity would be backfilled and sealed with natural materials. A Federal Environmental Assessment Review Panel was formed in 1992 to assess the concept for disposal of the spent fuel. In this paper a brief discussion of the human health impacts of the proposed concept is presented. Our assessment is based on the information provided by AECL, namely, the main EIS document, a summary and nine other supporting documents

Inspection of disposal canisters components

International Nuclear Information System (INIS)

Pitkaenen, J.

2013-12-01

This report presents the inspection techniques of disposal canister components. Manufacturing methods and a description of the defects related to different manufacturing methods are described briefly. The defect types form a basis for the design of non-destructive testing because the defect types, which occur in the inspected components, affect to choice of inspection methods. The canister components are to nodular cast iron insert, steel lid, lid screw, metal gasket, copper tube with integrated or separate bottom, and copper lid. The inspection of copper material is challenging due to the anisotropic properties of the material and local changes in the grain size of the copper material. The cast iron insert has some acoustical material property variation (attenuation, velocity changes, scattering properties), which make the ultrasonic inspection demanding from calibration point of view. Mainly three different methods are used for inspection. Ultrasonic testing technique is used for inspection of volume, eddy current technique, for copper components only, and visual testing technique are used for inspection of the surface and near surface area

Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

International Nuclear Information System (INIS)

Wiles, L.E.; McCann, R.A.

1981-09-01

This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

Background information for NDT qualification of Finnish disposal canisters of spent fuel

International Nuclear Information System (INIS)

Sarkimo, M.; Pitkaenen, J.

2013-12-01

This report presents a review to basic concepts, which are applied in the qualification of non-destructive testing (NDT) techniques. The qualification systems developed and used in some countries are briefly described in the beginning of the report. Anyway the report mainly discusses the qualification practices applied in the Finnish nuclear industry. The Finnish Radiation and Nuclear Safety Authority (STUK) in the YVL Guide 3.8 define the Finnish qualification approach applied for the in-service inspections. The principles presented in this document follow the views of the international organisations: Nuclear Regulator Working Group (NRWG) and European Network for Inspection and Qualification (ENIQ). For the practical qualification work a national guideline is established using so called SP-documents that include specific rules and instructions for execution of qualifications in accordance with YVL Guide 3.8 principles. Altogether the Finnish qualification system can be seen very well to follow the European (ENIQ) methodology. The report discusses several qualification terms and documents. Thus the normally necessary tasks and parts of a qualification are described. The qualification can be seen as a project that includes several tasks, which will be performed by different parties. Enough resources and time should be reserved for the planning and control of a qualification project to ensure its fluent progress. Some tasks are discussed in the report taking into account the situation in the qualification cases that are seen to be linked to the inspections of disposal canisters of spent fuel. (orig.)

Background information for NDT qualification of Finnish disposal canisters of spent fuel

Energy Technology Data Exchange (ETDEWEB)

Sarkimo, M. [VTT Technical Research Centre of Finland, Espoo (Finland); Pitkaenen, J.

2013-12-15

This report presents a review to basic concepts, which are applied in the qualification of non-destructive testing (NDT) techniques. The qualification systems developed and used in some countries are briefly described in the beginning of the report. Anyway the report mainly discusses the qualification practices applied in the Finnish nuclear industry. The Finnish Radiation and Nuclear Safety Authority (STUK) in the YVL Guide 3.8 define the Finnish qualification approach applied for the in-service inspections. The principles presented in this document follow the views of the international organisations: Nuclear Regulator Working Group (NRWG) and European Network for Inspection and Qualification (ENIQ). For the practical qualification work a national guideline is established using so called SP-documents that include specific rules and instructions for execution of qualifications in accordance with YVL Guide 3.8 principles. Altogether the Finnish qualification system can be seen very well to follow the European (ENIQ) methodology. The report discusses several qualification terms and documents. Thus the normally necessary tasks and parts of a qualification are described. The qualification can be seen as a project that includes several tasks, which will be performed by different parties. Enough resources and time should be reserved for the planning and control of a qualification project to ensure its fluent progress. Some tasks are discussed in the report taking into account the situation in the qualification cases that are seen to be linked to the inspections of disposal canisters of spent fuel. (orig.)

Fuel canister and blockage pin fabrication for SLSF Experiment P4

International Nuclear Information System (INIS)

Rhude, H.V.; Folkrod, J.R.; Noland, R.A.; Schaus, P.S.; Benecke, M.W.; Delucchi, T.A.

1983-01-01

As part of its fast breeder reactor safety research program, Argonne National Laboratory (ANL) has conducted an experiment (SLSF Experiment P4) to determine the extent of fuel-failure propagation resulting from the release of molten fuel from one or more heat-generating fuel canisters. The test conditions consisted of 37 full-length FTR fuel pins operating at FTR rated core nominal peak fuel/reduced coolant conditions. Thirty-four of the the fuel pins were prototypical FTR mixed-oxide fuel pins. The other three fuel pins were fabricated with a mid-core section having an enlarged canister containing fully enriched UO 2 . Two of the canisters were cylindrical and one was fluted. The cylindrical canisters were designed to fail and release molten fuel into the 37-pin fuel cluster at near full power

Interim report on safety assessment of spent fuel disposal TILA-96

Energy Technology Data Exchange (ETDEWEB)

Vieno, T.; Nordman, H. [VTT Energy, Espoo (Finland)

1996-12-01

The TILA-96 study, a continuation and update of the TVO-92 safety analysis for Finnish radioactive waste disposal, confirms that the planned system for spent fuel disposal fulfills the proposed safety criteria. Provided that no major disruptive event hits the repository, initially intact copper canisters preserve their integrity for millions of years and no significant amount of radioactive substances will ever escape from the repository. Impacts of potential canister failures have been analysed employing conservative assumptions, models and data. In the case of single canister failures, the results show that the margin to the proposed regulatory criteria is more than three orders of magnitude in the dose rate and more than four orders of magnitude in the release rates into the biosphere. Even in the extreme cases, where all 1500 canisters are assumed to be initially defective or to disappear simultaneously at 10 000 years in the worst possible location in the repository, all the proposed safety criteria would be passed. When realistic modelling and data are used in the consequence analyses, the results show negligible releases and doses. (refs.).

Interim report on safety assessment of spent fuel disposal TILA-96

International Nuclear Information System (INIS)

Vieno, T.; Nordman, H.

1996-12-01

The TILA-96 study, a continuation and update of the TVO-92 safety analysis for Finnish radioactive waste disposal, confirms that the planned system for spent fuel disposal fulfills the proposed safety criteria. Provided that no major disruptive event hits the repository, initially intact copper canisters preserve their integrity for millions of years and no significant amount of radioactive substances will ever escape from the repository. Impacts of potential canister failures have been analysed employing conservative assumptions, models and data. In the case of single canister failures, the results show that the margin to the proposed regulatory criteria is more than three orders of magnitude in the dose rate and more than four orders of magnitude in the release rates into the biosphere. Even in the extreme cases, where all 1500 canisters are assumed to be initially defective or to disappear simultaneously at 10 000 years in the worst possible location in the repository, all the proposed safety criteria would be passed. When realistic modelling and data are used in the consequence analyses, the results show negligible releases and doses. (refs.)

Canister materials proposed for final disposal of high level nuclear waste - a review with respect to corrosion resistance

Energy Technology Data Exchange (ETDEWEB)

Mattsson, E; Odoj, R; Merz, E [eds.

1981-06-01

Spent fuel from nuclear reactors has to be disposed of either after reprocessing or without such treatment. Due to toxic radiation the nuclear waste has to be isolated from the biosphere for 300-1000 years, or in extreme cases for more than 100,000 years. The nuclear waste will be enclosed in corrosion resistant canisters. These will be deposited in repositories in geological formations, such as granite, basalt, clay, bedded or domed salt, or the sediments beneath the deep ocean floor. There the canisters will be exposed to groundwater, brine or seawater at an elevated temperature. Species formed by radiolysis may affect the corrosivity of the agent. The corrosion resistance of candidate canister materials is evaluated by corrosion tests and by thermodynamic and mass transport calculations. Examination of ancient metal objects after long exposure in nature may give additional information. On the basis of the work carried out so far, the principal candidate canister materials are titanium materials, copper and high purity alumina.

Physical properties of encapsulate spent fuel in canisters

International Nuclear Information System (INIS)

1999-01-01

Spent fuel and high-level wastes will be permanently stored in a deep geological repository (AGP). Prior to this, they will be encapsulated in canisters. The present report is dedicated to the study of such canisters under the different physical demands that they may undergo, be those in operating or accident conditions. The physical demands of interest include mechanical demands, both static and dynamic, and thermal demands. Consideration is given to the complete file of the canister, from the time when it is empty and without lid to the final conditions expected in the repository. Thermal analyses of canisters containing spent fuel are often carried out in two dimensions, some times with hypotheses of axial symmetry and some times using a plane transverse section through the centre of the canister. The results obtained in both types of analyses are compared here to those of complete three-dimensional analyses. The latter generate more reliable information about the temperatures that may be experienced by the canister and its contents; they also allow calibrating the errors embodied in the two-dimensional calculations. (Author)

A study of the operational logistics in the disposal plant for spent nuclear fuel

International Nuclear Information System (INIS)

Sylvaenne, O.; Kaskinen, T.; Kuussaari, P.

2003-02-01

The final disposal plant for spent nuclear fuel comprises an encapsulation facility that will be built on the surface, other support activities above ground, and a repository that will be constructed deep in the bedrock. This report analyses the final repository operational logistics. The desktop research report is compiled of data taken from several existing planning reports covering the planning periods 1997-2002. The logistics specialised description of the final repository considers most areas in the daily operation of the facility. Among these are: Disposal tunnel excavation; construction and transports; Tunnel preparation for canisters; Reception of spent nuclear fuel transport casks; Encapsulation process; Preparation of bentonite blocks for canister holes, block laying; Final disposal of canisters; and Preparation of backfilling material and backfilling. The transport and handling volumes have certain cycles. Rock will be excavated during one contiguous period in 3 years, backfilling takes two weeks in a month and the deposition of canisters also two weeks. Thus the material flows vary greatly due to their cyclical nature. The transport and handling volumes are considerable, by far largest single item being excavated rock with about 5000 annual truck loads during the active excavation period, backfilling is about 1300 loads yearly at a steady pace. The report covers and summarises material flows, handling methods and equipment, buffering, storage and transports. It suggests some changes to operational procedures. Proposals have been made as to the location of the encapsulation facility and the methods of material transport. The logistical 'hot' issues, entry of the main transport ramp, rock field, rock crushing process, bentonite storage, bentonite brick production and backfiller production are all proposed to be located close to each other to minimise driving distance. It has also been proposed that the bentonite block buffer should rather be located at

Fuel and canister process report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Werme, Lars

2006-10-01

This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The report is written by, and for, experts in the relevant scientific fields. It should though be possible for a generalist in the area of long-term safety assessments of geologic nuclear waste repositories to comprehend the contents of the report. The report is an important part of the documentation of the SR-Can project and an essential reference within the project, providing a scientifically motivated plan for the handling of geosphere processes. It is, furthermore, foreseen that the report will be essential for reviewers scrutinising the handling of geosphere issues in the SR-Can assessment. Several types of fuel will be emplaced in the repository. For the reference case with 40 years of reactor operation, the fuel quantity from boiling water reactors, BWR fuel, is estimated at 7,000 tonnes, while the quantity from pressurized water reactors, PWR fuel, is estimated at about 2,300 tonnes. In addition, 23 tonnes of mixed-oxide fuel (MOX) fuel of German origin from BWR and PWR reactors and 20 tonnes of fuel from the decommissioned heavy water reactor in Aagesta will be disposed of. To allow for future changes in the Swedish nuclear programme, the safety assessment assumes a total of 6,000 canister corresponding to 12,000 tonnes of fuel

Fuel and canister process report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Werme, Lars (ed.)

2006-10-15

This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The report is written by, and for, experts in the relevant scientific fields. It should though be possible for a generalist in the area of long-term safety assessments of geologic nuclear waste repositories to comprehend the contents of the report. The report is an important part of the documentation of the SR-Can project and an essential reference within the project, providing a scientifically motivated plan for the handling of geosphere processes. It is, furthermore, foreseen that the report will be essential for reviewers scrutinising the handling of geosphere issues in the SR-Can assessment. Several types of fuel will be emplaced in the repository. For the reference case with 40 years of reactor operation, the fuel quantity from boiling water reactors, BWR fuel, is estimated at 7,000 tonnes, while the quantity from pressurized water reactors, PWR fuel, is estimated at about 2,300 tonnes. In addition, 23 tonnes of mixed-oxide fuel (MOX) fuel of German origin from BWR and PWR reactors and 20 tonnes of fuel from the decommissioned heavy water reactor in Aagesta will be disposed of. To allow for future changes in the Swedish nuclear programme, the safety assessment assumes a total of 6,000 canister corresponding to 12,000 tonnes of fuel.

A welding system for spent fuel canister lid

International Nuclear Information System (INIS)

Suikki, M.; Wendelin, T.

2008-06-01

The report presents a proposed welding system for spent fuel canister lids. The system is used for welding the copper lid to the copper overpack. The apparatus will be installed in the encapsulation plant. The report presents basic requirements for and implementation of the welding system, operation, service and maintenance of the equipment, as well as a cost estimate. Some aspects of the apparatus design are quite specified, but the actual detailed planning and final selection of components is not included. The report also describes actions for possible malfunction and fault conditions. Closing of the copper cylinder's lid is carried out by electron beam welding, which must be performed in vacuum. The welding system for spent fuel canister lid consists of two welding chambers, a canister docking system, an EB-welding machine with its accessories, a vacuum apparatus, as well as necessary auxiliary equipment. The system's equipment is housed in a welding room, an auxiliary system room, an operation control room, as well as mounted on the ceiling of a transfer corridor. One of the welding chambers is intended for carrying out test welding procedures and for calibration of welding parameters. The actual spent fuel canister lid welding chamber has a weldingready canister docked thereto in an airtight manner. The chamber is pumped for a vacuum, followed by closing the canister's copper lid and carrying out the lid welding process. The lid is brought into the chamber prior to docking the canister by means of a canister transfer trolley lifting gear. Lifting of the canister and rotating it during a welding process are also handled by means of the transfer trolley. The lid welding chamber houses equipment for the alignment and installation of the lid, as well as heating means for the top side of a copper overpack for ensuring a sufficient installation clearance between the lid and the overpack. The equipment not needed in the immediate vicinity of welding chambers, is

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

Prototype spent-fuel canister design, analysis, and test

International Nuclear Information System (INIS)

Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

1982-03-01

Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included

Reference concepts for the final disposal of LWR spent fuel and other high activity wastes in Spain

International Nuclear Information System (INIS)

Huertas, F.; Ulibarri, A.

1993-01-01

Studies over the last three years have been recently concluded with the selection of a reference repository concept for the final disposal of spent fuel and other high activity wastes in deep geological formations. Two non-site specific preliminary designs, at a conceptual level, have been developed; one considers granite as the host rock and the other rock salt formations. The Spanish General Radioactive Waste Program also considers clay as a potential host rock for HLW deep disposal; conceptualization for a deep repository in clay is in the initial phase of development. The salt repository concept contemplates the disposal of the HLW in self-shielding casks emplaced in the drifts of an underground facility, excavated at a depth of 850 m in a bedded salt formation. The Custos Type I(7) cask admits up to seven intact PWR fuel assemblies or 21 of BWR type. The final repository facilities are planned to accept a total of 20,000 fuel assemblies (PWR and BWR) and 50 vitrified waste canisters over a period of 25 years. The total space needed for the surface facilities amounts to 322,000 m 2 , including the rock salt dump. The space required for the underground facilities amounts to 1.2 km 2 , approximately. The granite repository concept contemplates the disposal of the HLW in carbon steel canisters, embedded in a 0.75 m thick buffer of swelling smectite clay, in the drifts of an underground facility, excavated at a depth of 55 m in granite. Each canister can host 3 PWR or 9 BWR fuel assemblies. For this concept the total number of canisters needed amounts to 4,860. The space required for the surface and underground facilities is similar to that of the salt concept. The technical principles and criteria used for the design are discussed, and a description of the repository concept is presented

TMI-2 fuel canister interface requirements for INEL. Revision 1

International Nuclear Information System (INIS)

Wilkins, D.E.; Martz, D.E.; Reno, H.W.

1984-06-01

This report focuses on fuel canister interface requirements at INEL which should be incorporated into the canister design criteria. The requirements will ensure compatibility with existing INEL structures and equipment to be used for receipt, unloading, and storage of fuel canisters. INEL can and does receive and store radioactive materials in many different forms, including reactor fuel. INEL requires detailed descriptions of canisters and casks. Therefore, requirements listed represent engineering design features which will simplify the handling and storage operations; consequently, they are not to be viewed as absolute or non-negotiable. However, the core acquisition contract was negotiated with certain storage assumptions which effect costs of storage. Deviations from those assumptions which significantly effect costs would require approval by DOE-Idaho. If some stated requirements are too restrictive, modifications based on sound engineering principles may be negotiated with INEL. 11 figures

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

Long-Term Dry Storage of High Burn-Up Spent Pressurized Water Reactor (PWR) Fuel in TAD (Transportation, Aging, and Disposal) Containers

International Nuclear Information System (INIS)

Hwang, Yong Soo

2008-12-01

A TAD canister, in conjunction with specially-designed over-packs can accomplish the functions of transportation, aging, and disposal (TAD) in the management of spent nuclear fuel (SNF). Industrial dry cask systems currently available for SNF are licensed for storage-only or for dual-purpose (i.e., storage and transportation). By extending the function to include the indefinite storage and perhaps, eventual geologic disposal, the TAD canister would have to be designed to enhance, among others, corrosion resistance, thermal stability, and criticality-safety control. This investigative paper introduces the use of these advanced iron-based, corrosion-resistant materials for SNF transportation, aging, and disposal.The objective of this investigative project is to explore the interest that KAERI would research and develop its specific SAM coating materials for the TAD canisters to satisfy the requirements of corrosion-resistance, thermal stability, and criticality-controls for long-term dry storage of high burn-up spent PWR fuel

Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

International Nuclear Information System (INIS)

HEARD, F.J.

1999-01-01

A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels

Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

Energy Technology Data Exchange (ETDEWEB)

HEARD, F.J.

1999-04-09

A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

Corrosion resistance of metal materials for HLW canister

International Nuclear Information System (INIS)

Furuya, Takashi; Muraoka, Susumu; Tashiro, Shingo

1982-02-01

In order to verify the materials as an important artificial barrier for canister of vitrified high-level waste from spent fuel reprocessing, data and reports were researched on corrosion resistance of the materials under conditions from glass form production to final disposal. Then, in this report, investigated subjects, improvement methods and future subjects are reviewed. It has become clear that there would be no problem on the inside and outside corrosion of the canister during glass production, but long term corrosion and radiation effect tests and the vitrification methods would be subjects in future on interim storage and final disposal conditions. (author)

Shippingport Spent Fuel Canister (SSFC) Design Report Project W-518

International Nuclear Information System (INIS)

JOHNSON, D.M.

2000-01-01

The SSFC Design Report Describes A spent fuel canister for Shippingport Core 2 blanket fuel assemblies. The design of the SSFC is a minor modification of the MCO. The modification is limited to the Shield Plug which remains unchanged with regard to interfaces with the canister shell. The performance characteristics remain those for the MCO, which bounds the payload of the SSFC

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)

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.

Storage and disposal of radioactive waste as glass in canisters

International Nuclear Information System (INIS)

Mendel, J.E.

1978-12-01

A review of the use of waste glass for the immobilization of high-level radioactive waste glass is presented. Typical properties of the canisters used to contain the glass, and the waste glass, are described. Those properties are used to project the stability of canisterized waste glass through interim storage, transportation, and geologic disposal

CANISTER TRANSFER SYSTEM DESCRIPTION DOCUMENT

International Nuclear Information System (INIS)

B. Gorpani

2000-01-01

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

Canister Transfer System Description Document

International Nuclear Information System (INIS)

2000-01-01

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

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

Survey of waste package designs for disposal of high-level waste/spent fuel in selected foreign countries

International Nuclear Information System (INIS)

Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

1989-09-01

This report presents the results of a survey of the waste package strategies for seven western countries with active nuclear power programs that are pursuing disposal of spent nuclear fuel or high-level wastes in deep geologic rock formations. Information, current as of January 1989, is given on the leading waste package concepts for Belgium, Canada, France, Federal Republic of Germany, Sweden, Switzerland, and the United Kingdom. All but two of the countries surveyed (France and the UK) have developed design concepts for their repositories, but none of the countries has developed its final waste repository or package concept. Waste package concepts are under study in all the countries surveyed, except the UK. Most of the countries have not yet developed a reference concept and are considering several concepts. Most of the information presented in this report is for the current reference or leading concepts. All canisters for the wastes are cylindrical, and are made of metal (stainless steel, mild steel, titanium, or copper). The canister concepts have relatively thin walls, except those for spent fuel in Sweden and Germany. Diagrams are presented for the reference or leading concepts for canisters for the countries surveyed. The expected lifetimes of the conceptual canisters in their respective disposal environment are typically 500 to 1,000 years, with Sweden's copper canister expected to last as long as one million years. Overpack containers that would contain the canisters are being considered in some of the countries. All of the countries surveyed, except one (Germany) are currently planning to utilize a buffer material (typically bentonite) surrounding the disposal package in the repository. Most of the countries surveyed plan to limit the maximum temperature in the buffer material to about 100 degree C. 52 refs., 9 figs

Source term for the bounding assessment of the Canadian nuclear fuel waste disposal concept

International Nuclear Information System (INIS)

Flavelle, P.

1996-02-01

This is the second in a series to derive the bounds of the post-closure hazard of the Canadian nuclear fuel waste disposal concept, based on the premise that it is unnecessary to predict accurately the real hazard if the bounding hazard can be shown to be acceptable. In this report a reference used (Bruce A fuel, 865 GJ/kgU average burnup) is used to derive the source term for contaminant releases from the emplacement canisters. This requires development of a container failure function which defines the age of the fuel when the canister is perforated and flooded. The source term is expressed as the time-dependent fractional release rate from the used fuel or as the time-dependent contaminant concentrations in the canister porewater. It is derived as the superposition of an instant release, comprising the upper bound of the gap and grain boundary inventory in the used fuel, and the long-term dissolution of the used fuel matrix. Several dissolution models (stoichiometric dissolution/preferential leaching) under different conditions (matrix solubility limited/ unlimited; oxidizing/ reducing solubility limits; groundwater flow/ no flow) are evaluated and the one resulting in the highest release rate/ highest porewater concentration is adopted as the bounding case. Comparisons between the models are made on the basis of the potential ingestion hazard of the canister porewater, to account for differences in the hazard of different radionuclides. (author) 20 refs., 4 tabs., 9 figs

Defense High Level Waste Disposal Container System Description Document

International Nuclear Information System (INIS)

Pettit, N. E.

2001-01-01

The Defense High Level Waste 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 and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms [IPWF]) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. US Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as co-disposal. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister inserted in the center and/or one or more DOE SNF canisters displacing a HLW canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by

Critical review of welding technology for canisters for disposal of spent fuel and high level waste

International Nuclear Information System (INIS)

Pike, S.; Allen, C.; Punshon, C.; Threadgill, P.; Gallegillo, M.; Holmes, B.; Nicholas, J.

2010-03-01

Nagra is the Swiss national cooperative for the disposal of radioactive waste and is responsible for final disposal of all types of waste produced in Switzerland, which are partitioned into two repository types, one for spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate level waste and one for low and intermediate level waste. In the general licences applied for these repositories, documentation has to show that long-term safety can be ensured and that factors for the construction, operation, and closure of the facility have been considered. Nagra has commissioned TWI to carry out a critical review of welding technologies for the sealing of HLW and SF canisters made of carbon steel. In conjunction with a material selection report, the information gained will be used as a preliminary step to provide input to developing design concepts for the canisters. The features to be considered are: a) Suitability of techniques for thickness of weld required; b) Suitability for remote operation, maintenance and set-up; c) Welding speed, weld quality, tolerances and cost; d) Effect of welding process on parent materials properties including microstructure corrosion resistance, distortion and residual stress; e) Potential post-weld treatments to reduce residual stress and enhance corrosion resistance; f) Suitability of inspection techniques for the weld thickness required; g) Impact of welding techniques on the canister design and material selection; h) Critique of emerging technologies which may be suitable in the future. The review of potential welding technologies began with a feasibility study carried out by TWI experts, where the unsuitable processes were rejected. For the remaining processes attention was focused on previous applications for the material and thickness suggested, and especially on safety critical applications such as applied in the nuclear and pressure vessel industry. Once the relevant information was gathered each process was

Disposal facility in Olkiluoto, description of above ground facilities in tunnel transport alternative

International Nuclear Information System (INIS)

Kukkola, T.

2006-11-01

The above ground facilities of the disposal plant on the Olkiluoto site are described in this report as they will be when the operation of the disposal facility starts in the year 2020. The disposal plant is visualised on the Olkiluoto site. Parallel construction of the deposition tunnels and disposal of the spent fuel canisters constitute the principal design basis of the disposal plant. The annual production of disposal canisters for spent fuel amounts to about 40. Production of 100 disposal canisters has been used as the capacity basis. Fuel from the Olkiluoto plant and from the Loviisa plant will be encapsulated in the same production line. The disposal plant will require an area of about 15 to 20 hectares above ground level. The total building volume of the above ground facilities is about 75000 m 3 . The purpose of the report is to provide the base for detailed design of the encapsulation plant and the repository spaces, as well as for coordination between the disposal plant and ONKALO. The dimensioning bases for the disposal plant are shown in the Tables at the end of the report. The report can also be used as a basis for comparison in deciding whether the fuel canisters are transported to the repository by a lift or a by vehicle along the access tunnel. (orig.)

Disposal facility in olkiluoto, description of above ground facilities in lift transport alternative

International Nuclear Information System (INIS)

Kukkola, T.

2006-11-01

The above ground facilities of the disposal plant on the Olkiluoto site are described in this report as they will be when the operation of the disposal facility starts in the year 2020. The disposal plant is visualised on the Olkiluoto site. Parallel construction of the deposition tunnels and disposal of the spent fuel canisters constitute the principal design basis of the disposal plant. The annual production of disposal canisters for spent fuel amounts to about 40. Production of 100 disposal canisters has been used as the capacity basis. Fuel from the Olkiluoto plant and from the Loviisa plant will be encapsulated in the same production line. The disposal plant will require an area of about 15 to 20 hectares above ground level. The total building volume of the above ground facilities is about 75000 m 3 . The purpose of the report is to provide the base for detailed design of the encapsulation plant and the repository spaces, as well as for coordination between the disposal plant and ONKALO. The dimensioning bases for the disposal plant are shown in the Tables at the end of the report. The report can also be used as a basis for comparison in deciding whether the fuel canisters are transported to the repository by a lift or by a vehicle along the access tunnel. (orig.)

Disposal of spent fuel from German nuclear power plants - paper work or technology?

International Nuclear Information System (INIS)

Graf, R.; Filbert, W.

2006-01-01

The reference concept 'direct disposal of spent fuel' was developed as an alternative to spent fuel reprocessing and vitrified HLW disposal. The technical facilities necessary for the implementation of this reference concept - the so called POLLUX-concept, e.g. interim storages 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 repository were tested aboveground in a test facility at a 1:1 scale. To optimise the concept all operational steps are reviewed for possible improvement. Most promising are a concept using canisters (BSK 3) instead of POLLUX casks, and the direct disposal of transport and storage casks (DIREGT-concept) which is the most recent one and has been designed for the direct disposal of large transport and storage casks. The final exploration of the pre-selected repository site is still pending, from the industries point of view due to political reasons only. The present paper describes the main concepts and their status as of today. (author)

Materials for Consideration in Standardized Canister Design Activities.

Energy Technology Data Exchange (ETDEWEB)

Bryan, Charles R.; Ilgen, Anastasia Gennadyevna; Enos, David George; Teich-McGoldrick, Stephanie; Hardin, Ernest

2014-10-01

This document identifies materials and material mitigation processes that might be used in new designs for standardized canisters for storage, transportation, and disposal of spent nuclear fuel. It also addresses potential corrosion issues with existing dual-purpose canisters (DPCs) that could be addressed in new canister designs. The major potential corrosion risk during storage is stress corrosion cracking of the weld regions on the 304 SS/316 SS canister shell due to deliquescence of chloride salts on the surface. Two approaches are proposed to alleviate this potential risk. First, the existing canister materials (304 and 316 SS) could be used, but the welds mitigated to relieve residual stresses and/or sensitization. Alternatively, more corrosion-resistant steels such as super-austenitic or duplex stainless steels, could be used. Experimental testing is needed to verify that these alternatives would successfully reduce the risk of stress corrosion cracking during fuel storage. For disposal in a geologic repository, the canister will be enclosed in a corrosion-resistant or corrosion-allowance overpack that will provide barrier capability and mechanical strength. The canister shell will no longer have a barrier function and its containment integrity can be ignored. The basket and neutron absorbers within the canister have the important role of limiting the possibility of post-closure criticality. The time period for corrosion is much longer in the post-closure period, and one major unanswered question is whether the basket materials will corrode slowly enough to maintain structural integrity for at least 10,000 years. Whereas there is extensive literature on stainless steels, this evaluation recommends testing of 304 and 316 SS, and more corrosion-resistant steels such as super-austenitic, duplex, and super-duplex stainless steels, at repository-relevant physical and chemical conditions. Both general and localized corrosion testing methods would be used to

Near-field performance of the advanced cold process canister

International Nuclear Information System (INIS)

Werme, L.

1990-09-01

A near-field performance evaluation of an Advanced Cold Process Canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. The canister design was originally proposed by TVO. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. Throughout the analysis, present day underground conditions has been assumed to persist during the service life of the canister. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie localized corrosion for the steel or copper canisters can be dismissed as a failure mechanism. The evaluation of the effects of processes outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. This factor will ensure the safety of the concept. (orig.)

Mechanical analysis of cylindrical part of canisters for spent nuclear fuel

International Nuclear Information System (INIS)

Ikonen, K.

2005-06-01

This report describes mechanical analyses of cylindrical part of the VVER 440-, BWR and EPR-type canisters for spent nuclear fuel. The task was first to evaluate the stresses at maximum design pressure and further by increasing pressure load to determine the limit collapse load and corresponding safety factor. Maximum design pressure 44 MPa is a sum of the hydrostatic pressure 30 MPa caused by 3 km ice layer, 7 MPa caused by ground water pressure at the deepest disposal depth of 700 m and 7 MPa from bentonite swelling pressure. The analysis presented in this report concern the middle area of the canisters, where the cast iron insert is considered to be more critical than in the ends of the canister. For the model a piece from the middle area of the canister was separated by two planes perpendicular to the axis of the canister. This piece was studied first by two-dimensional plane strain model, where the planes are constrained and no elongation of the canister takes place. In the second model one of the planes was constrained and the other plane was allowed to displace in axial direction, which remains as a plane during deformation and to which axial pressure force is directed. This analysis, which corresponds better the real condition in the canister, was performed as threedimensional. The analyses gave however practically equal results due to plastic deformation. Thus the analysis can be done by two-dimensional plane strain model leading to same accuracy with less computation effort. Analyses were performed as large displacement and large strain analyses by the PASULA computing package, which has been developed at VTT for a variety of structural analysis and for heat conduction calculations. A special routine was developed for automatic mesh generation. Before the analysis of the VVER 440-, BWR- and EPR-type canisters the calculation methodology was validated with test results, which were received from pressure tests performed with a short BWR canister in Germany

Groundwork for Universal Canister System Development

Energy Technology Data Exchange (ETDEWEB)

Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gross, Mike [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Prouty, Jeralyn L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rigali, Mark J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Craig, Brian [Argonne National Lab. (ANL), Argonne, IL (United States); Han, Zenghu [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, John Hok [Argonne National Lab. (ANL), Argonne, IL (United States); Liu, Yung [Argonne National Lab. (ANL), Argonne, IL (United States); Pope, Ron [Argonne National Lab. (ANL), Argonne, IL (United States); Connolly, Kevin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Feldman, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jarrell, Josh [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Radulescu, Georgeta [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wells, Alan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-01

The mission of the United States Department of Energy's Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and go vernment - sponsored nuclear energy re search. S ome of the waste s that that must be managed have be en identified as good candidates for disposal in a deep borehole in crystalline rock (SNL 2014 a). In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister - based system that can be used for handling these wastes during the disposition process (i.e., storage, transfers, transportation, and disposal) could facilitate the eventual disposal of these wastes. This report provides information for a program plan for developing specifications regarding a canister - based system that facilitates small waste form packaging and disposal and that is integrated with the overall efforts of the DOE's Office of Nuclear Energy Used Fuel Dis position Camp aign's Deep Borehole Field Test . Groundwork for Universal Ca nister System Development September 2015 ii W astes to be considered as candidates for the universal canister system include capsules containing cesium and strontium currently stored in pools at the Hanford Site, cesium to be processed using elutable or nonelutable resins at the Hanford Site, and calcine waste from Idaho National Laboratory. The initial emphasis will be on disposal of the cesium and strontium capsules in a deep borehole that has been drilled into crystalline rock. Specifications for a universal canister system are derived from operational, performance, and regulatory requirements for storage, transfers, transportation, and disposal of radioactive waste. Agreements between the Department of Energy and the States of Washington and Idaho, as well as the Deep Borehole Field Test plan provide schedule requirements for development of the universal canister system

Production methods and costs of oxygen free copper canisters for nuclear waste disposal

International Nuclear Information System (INIS)

Rajainmaeki, H.; Nieminen, M.; Laakso, L.

1991-08-01

The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for spent fuel repository are discussed. The capsule design is based on the TVO's new advanced cold process concept where a steel canister is surrounded by the oxygen free copper canister. This study shows that already at present there exist several possible manufacturing routes, which result in consistently high quality canisters. Hot rolling, bending and EB-welding the seam is the best way to assure the small grain size which is preferable for the best inspectability of the final EB-welded seam of the lid. The same route turns out also to be the most economical

Production methods and costs of oxygen free copper canisters for nuclear waste disposal

International Nuclear Information System (INIS)

Rajainmaeki, H.; Nieminen, M.; Laakso, L.

1991-06-01

The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for spent fuel repository are discussed. The capsule design is based on the TVO's new advanced cold process concept where a steel canister is surrounded by the oxygen free copper canister. This study shows that already at present there exist several possible manufacturing routes, which results in consistently high quality canisters. Hot rolling, bending and EB-welding the seam is the best way to assure the small grain size which is preferable for the best inspectability of the final EB-welded seam of the lid. The same route turns out also to be the most economical. (au)

Choices of canisters and elements for the first fuel shipment from K West Basin

International Nuclear Information System (INIS)

Makenas, B.J.

1995-03-01

Twenty-two canisters (10 prime and 12 backup candidates) in the K West Basin have been identified as containing fuel which, when examined, will satisfy the Data Quality Objectives for the first fuel shipment from this basin. These were chosen as meeting criteria such as containing relatively long fuel elements, locking bar integrity, and the availability of gas/liquid interface level measurements for associated canister gas traps. Two canisters were identified as having reported broken fuel on initial loading. Usage and interpretation of canister cesium concentration measurements have also been established and levels of maximum and minimum acceptable cesium concentration (from a data optimization point of view) for decapping have been determined although other operational cesium limits may also apply. Criteria for picking particular elements, once a canister is opened, are reviewed in this document. A pristine, a slightly damaged, and a badly damaged element are desired. The latter includes elements with end caps removed but does not include elements which have large amounts of swelling or split cladding that might interfere with handling tools. Finally, operational scenarios have been suggested to aid in the selections of canisters and elements in a way that utilizes anticipated canister gas sampling and leads to a correct and quick choice of elements which will supply the desired data

Life cycle assessment of geological repositories for the final disposal of spent fuel in Finland and Sweden

International Nuclear Information System (INIS)

Puhrer, A.; Bauer, C.

2014-01-01

This paper presents a Life Cycle Assessment (LCA) of the geological repositories for the final disposal of spent nuclear fuel in Finland and Sweden. A separate LCA has been performed for the geological spent fuel repository in each country and the results have been compared. A further benchmark comparison has been made with the LCA of the Swiss geological repository for high-level waste and spent fuel. The life cycle inventory (LCI) product system boundaries include the spent fuel repository and encapsulation facility in each country. All materials, processes, consumed utilities and transport associated with the construction, operation and closure of the repositories for spent fuel are included in the LCI. The life cycle impact assessment (LCIA) is performed using two methods: IPCC 2007 Climate Change and ReCiPe. These assessment methods return results pertaining to global warming potential (GWP) as well as a number of environmental impact categories such as human toxicity and natural land transformation. Results indicate that the use of copper for disposal canister fabrication and bentonite for repository backfilling are the causes for most of the environmental impact of the spent fuel repositories in Finland and Sweden. Alternate, less bentonite-intensive backfilling scenarios may mitigate this impact. While the Swiss bentonite consumption is lower and no copper is used for canister fabrication, the Swiss electricity and fuel consumption associated with final disposal of high-level waste and spent fuel is significantly higher than in Finland or Sweden. Approximately 1 g CO 2 -eq is emitted due to the final disposal of spent fuel and HLW per kWh of nuclear generated electricity. This represents some 10% of the emissions due to the entire nuclear energy chain and is practically negligible in the context of GHG emissions of other energy technologies. (authors)

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Spent nuclear fuel canister storage building conceptual design report

Energy Technology Data Exchange (ETDEWEB)

Swenson, C.E. [Westinghouse Hanford Co., Richland, WA (United States)

1996-01-01

This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ``Technical Baseline and Updated Cost Estimate for the Canister Storage Building``, dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995.

Spent nuclear fuel canister storage building conceptual design report

International Nuclear Information System (INIS)

Swenson, C.E.

1996-01-01

This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ''Technical Baseline and Updated Cost Estimate for the Canister Storage Building'', dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995

Disposability Assessment: Aluminum-Based Spent Nuclear Fuel Forms

Energy Technology Data Exchange (ETDEWEB)

Vinson, D.W.

1998-11-06

This report provides a technical assessment of the Melt-Dilute and Direct Al-SNF forms in disposable canisters with respect to meeting the requirements for disposal in the Mined Geologic Disposal System (MGDS) and for interim dry storage in the Treatment and Storage Facility (TSF) at SRS.

White Paper: Multi-purpose canister (MPC) for DOE-owned spent nuclear fuel (SNF)

International Nuclear Information System (INIS)

Knecht, D.A.

1994-04-01

The paper examines the issue, What are the advantages, disadvantages, and other considerations for using the MPC concept as part of the strategy for interim storage and disposal of DOE-owned SNF? The paper is based in part on the results of an evaluation made for the DOE National Spent Fuel Program by the Waste Form Barrier/Canister Team, which is composed of knowledgeable DOE and DOE-contractor personnel. The paper reviews the MPC and DOE SNF status, provides criteria and other considerations applicable to the issue, and presents an evaluation, conclusions, and recommendations. The primary conclusion is that while most of DOE SNF is not currently sufficiently characterized to be sealed into an MPC, the advantages of standardized packages in handling, reduced radiation exposure, and improved human factors should be considered in DOE SNF program planning. While the design of MPCs for DOE SNF are likely premature at this time, the use of canisters should be considered which are consistent with interim storage options and the MPC design envelope

The concrete canister program

International Nuclear Information System (INIS)

Ohta, M.M.

1978-02-01

In the spring of 1974, WNRE began development and demonstration of a dry storage concept, called the concrete canister, as a possible alternative to storage of irradiated CANDU fuel in water pools. The canister is a thick-walled concrete monolith containing baskets of fuel in the dry state. The decay heat from the fuel is dissipated to the environment by natural heat transfer. Four canisters were designed and constructed. Two canisters containing electric heaters have been subjected to heat loads of 2.5 times the design, ramp heat-load cycling, and simulated weathering tests. The other two canisters were loaded with irradiated fuel, one containing fuel bundles of uniform decay heat and the other containing bundles of non-uniform decay heat in a non-symmetrical radial and axial array. The collected data were used to verify the analytical tools for prediction of effectiveness of heat transfer and radiation shielding and to verify the design of the basket and canisters. The demonstration canisters have shown that this concept is a viable alternative to water pools for the storage of irradiated CANDU fuel. (author)

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.

Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Ryskamp, J.M.; Adams, J.P.; Faw, E.M.; Anderson, P.A.

1996-09-01

Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments.

Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

International Nuclear Information System (INIS)

Ryskamp, J.M.; Adams, J.P.; Faw, E.M.; Anderson, P.A.

1996-09-01

Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments

Thermal analysis of dry concrete canister storage system for CANDU spent fuel

International Nuclear Information System (INIS)

Ryu, Yong Ho

1992-02-01

This paper presents the results of a thermal analysis of the concrete canisters for interim dry storage of spent, irradiated Canadian Deuterium Uranium(CANDU) fuel. The canisters are designed to contain 6-year-old fuel safely for periods of 50 years in stainless steel baskets sealed inside a steel-lined concrete shield. In order to assure fuel integrity during the storage, fuel rod temperature shall not exceed the temperature limit. The contents of thermal analysis include the following : 1) Steady state temperature distributions under the conservative ambient temperature and insolation load. 2) Transient temperature distributions under the changes in ambient temperature and insolation load. Accounting for the coupled heat transfer modes of conduction, convection, and radiation, the computer code HEATING5 was used to predict the thermal response of the canister storage system. As HEATING5 does not have the modeling capability to compute radiation heat transfer on a rod-to-rod basis, a separate calculating routine was developed and applied to predict temperature distribution in a fuel bundle. Thermal behavior of the canister is characterized by the large thermal mass of the concrete and radiative heat transfer within the basket. The calculated results for the worst case (steady state with maximum ambient temperature and design insolation load) indicated that the maximum temperature of the 6 year cooled fuel reached to 182.4 .deg. C, slightly above the temperature limit of 180 .deg. C. However,the thermal inertia of the thick concrete wall moderates the internal changes and prevents a rise in fuel temperature in response to ambient changes. The maximum extent of the transient zone was less than 75% of the concrete wall thickness for cyclic insolation changes. When transient nature of ambient temperature and insolation load are considered, the fuel temperature will be a function of the long term ambient temperature as opposed to daily extremes. The worst design

Safety assessment of spent fuel disposal in Haestholmen, Kivetty, Olkiluoto and Romuvaara - TILA-99

Energy Technology Data Exchange (ETDEWEB)

Vieno, T.; Nordman, H. [VTT Energy (Finland)

1999-03-01

The spent fuel from the Finnish nuclear power plants is planned to be disposed of in copper-iron canisters emplaced in a KBS-3 type repository constructed at a depth of about 500 metres at one of the four candidate sites investigated. The disposal concept aims at long-term isolation of the spent fuel assemblies from the biosphere and even from the geosphere. The evaluation of the normal evolution of the disposal system accords with the conclusions of the previous Finnish, Swedish and Canadian safety assessments of similar disposal concepts. Subject to the influence of the expected, normal evolution of the repository, initially intact copper-iron canisters will most likely preserve their integrity for more than one million years at any of the candidate sites. Consequently, the best-estimate assessment is that there never will be any significant releases of radionuclides from the repository into the geosphere. Consequences of potential canister failures have been evaluated using conservative assumptions, models and data. The results show that at any of the sites a large number of canisters could be assumed to be initially defective or to `disappear` simultaneously after some time without that the proposed constraints for release rates into the biosphere or dose rates were exceeded. In most cases this conclusion is valid for all canisters failing simultaneously, even if rather pessimistic flow and transport data is used. In the sensitivity and `what if` analyses where very high flow rates of saline groundwater are assumed, highest release and dose rates are caused by weakly-sorbing cations Sr-90 and Ra-226. The most important differences between the sites are related to the coastal location and brackish/saline groundwater of Haestholmen and Olkiluoto, and on the other hand to the inland location and fresh groundwater of Kivetty and Romuvaara. Because of the ongoing postglacial land uplift at the coast of the Baltic Sea, Olkiluoto and Haestholmen, too, may become

Safety assessment of spent fuel disposal in Haestholmen, Kivetty, Olkiluoto and Romuvaara - TILA-99

International Nuclear Information System (INIS)

Vieno, T.; Nordman, H.

1999-03-01

The spent fuel from the Finnish nuclear power plants is planned to be disposed of in copper-iron canisters emplaced in a KBS-3 type repository constructed at a depth of about 500 metres at one of the four candidate sites investigated. The disposal concept aims at long-term isolation of the spent fuel assemblies from the biosphere and even from the geosphere. The evaluation of the normal evolution of the disposal system accords with the conclusions of the previous Finnish, Swedish and Canadian safety assessments of similar disposal concepts. Subject to the influence of the expected, normal evolution of the repository, initially intact copper-iron canisters will most likely preserve their integrity for more than one million years at any of the candidate sites. Consequently, the best-estimate assessment is that there never will be any significant releases of radionuclides from the repository into the geosphere. Consequences of potential canister failures have been evaluated using conservative assumptions, models and data. The results show that at any of the sites a large number of canisters could be assumed to be initially defective or to 'disappear' simultaneously after some time without that the proposed constraints for release rates into the biosphere or dose rates were exceeded. In most cases this conclusion is valid for all canisters failing simultaneously, even if rather pessimistic flow and transport data is used. In the sensitivity and 'what if' analyses where very high flow rates of saline groundwater are assumed, highest release and dose rates are caused by weakly-sorbing cations Sr-90 and Ra-226. The most important differences between the sites are related to the coastal location and brackish/saline groundwater of Haestholmen and Olkiluoto, and on the other hand to the inland location and fresh groundwater of Kivetty and Romuvaara. Because of the ongoing postglacial land uplift at the coast of the Baltic Sea, Olkiluoto and Haestholmen, too, may become

Remote Welding, NDE and Repair of DOE Standardized Canisters

Energy Technology Data Exchange (ETDEWEB)

Eric Larsen; Art Watkins; Timothy R. McJunkin; Dave Pace; Rodney Bitsoi

2006-05-01

The U.S. Department of Energy (DOE) created the National Spent Nuclear Fuel Program (NSNFP) to manage DOE’s spent nuclear fuel (SNF). One of the NSNFP’s tasks is to prepare spent nuclear fuel for storage, transportation, and disposal at the national repository. As part of this effort, the NSNFP developed a standardized canister for interim storage and transportation of SNF. These canisters will be built and sealed to American Society of Mechanical Engineers (ASME) Section III, Division 3 requirements. Packaging SNF usually is a three-step process: canister loading, closure welding, and closure weld verification. After loading SNF into the canisters, the canisters must be seal welded and the welds verified using a combination of visual, surface eddy current, and ultrasonic inspection or examination techniques. If unacceptable defects in the weld are detected, the defective sections of weld must be removed, re-welded, and re-inspected. Due to the high contamination and/or radiation fields involved with this process, all of these functions must be performed remotely in a hot cell. The prototype apparatus to perform these functions is a floor-mounted carousel that encircles the loaded canister; three stations perform the functions of welding, inspecting, and repairing the seal welds. A welding operator monitors and controls these functions remotely via a workstation located outside the hot cell. The discussion describes the hardware and software that have been developed and the results of testing that has been done to date.

Analysis of probability of defects in the disposal canisters

International Nuclear Information System (INIS)

Holmberg, J.-E.; Kuusela, P.

2011-06-01

This report presents a probability model for the reliability of the spent nuclear waste final disposal canister. Reliability means here that the welding of the canister lid has no critical defects from the long-term safety point of view. From the reliability point of view, both the reliability of the welding process (that no critical defects will be born) and the non-destructive testing (NDT) process (all critical defects will be detected) are equally important. In the probability model, critical defects in a weld were simplified into a few types. Also the possibility of human errors in the NDT process was taken into account in a simple manner. At this moment there is very little representative data to determine the reliability of welding and also the data on NDT is not well suited for the needs of this study. Therefore calculations presented here are based on expert judgements and on several assumptions that have not been verified yet. The Bayesian probability model shows the importance of the uncertainty in the estimation of the reliability parameters. The effect of uncertainty is that the probability distribution of the number of defective canisters becomes flat for larger numbers of canisters compared to the binomial probability distribution in case of known parameter values. In order to reduce the uncertainty, more information is needed from both the reliability of the welding and NDT processes. It would also be important to analyse the role of human factors in these processes since their role is not reflected in typical test data which is used to estimate 'normal process variation'.The reported model should be seen as a tool to quantify the roles of different methods and procedures in the weld inspection process. (orig.)

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.

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

Modelling studies for the assessment of the Advanced Cold Process Canister

International Nuclear Information System (INIS)

Henshaw, J.; Hoch, A.R.; Sharland, S.M.

1991-01-01

The Advanced Cold Process Canister (ACPC) is a new concept for the encapsulation of spent nuclear fuel for geological disposal. It consists of steel canister encased in a copper overpack. In this paper, modelling studies to assess the performance of the ACPC under repository conditions are presented. The production of nitric acid and ammonia through radiolysis of any water remaining inside the canister under fault conditions has been examined in this study. However, results suggest that only low levels are possible, and the risk of stress-corrosion cracking is considered small. The corrosion behavior subsequent to a breach in the outer canister was also considered. A model was constructed to predict the hydrogen gas production due to corrosion reactions, and evolution of the corrosion behavior

Encapsulation and handling of spent nuclear fuel for final disposal

International Nuclear Information System (INIS)

Loennerberg, B.; Larker, H.; Ageskog, L.

1983-05-01

The handling and embedding of those metal parts which arrive to the encapsulation station with the fuel is described. For the encapsulation of fuel two alternatives are presented, both with copper canisters but with filling of lead and copper powder respectively. The sealing method in the first case is electron beam welding, in the second case hot isostatic pressing. This has given the headline of the two chapters describing the methods: Welded copper canister and Pressed copper canister. Chapter 1, Welded copper canister, presents the handling of the fuel when it arrives to the encapsulation station, where it is first placed in a buffer pool. From this pool the fuel is transferred to the encapsulation process and thereby separated from fuel boxes and boron glass rod bundles, which are transported together with the fuel. The encapsulation process comprises charging into a copper canister, filling with molten lead, electron beam welding of the lid and final inspection. The transport to and handling in the final repository are described up to the deposition and sealing in the deposition hole. Handling of fuel residues is treated in one of the sections. In chapter 2, Pressed copper canister, only those parts of the handling, which differ from chapter 1 are described. The hot isostatic pressing process is given in the first sections. The handling includes drying, charging into the canister, filling with copper powder, seal lid application and hot isostatic pressing before the final inspection and deposition. In the third chapter, BWR boxes in concrete moulds, the handling of the metal parts, separated from the fuel, are dealt with. After being lifted from the buffer pool they are inserted in a concrete mould, the mould is filled with concrete, covered with a lid and after hardening transferred to its own repository. The deposition in this repository is described. (author)

Drying tests conducted on Three Mile Island fuel canisters containing simulated debris

International Nuclear Information System (INIS)

Palmer, A.J.

1995-01-01

Drying tests were conducted on TMI-2 fuel canisters filled with simulated core debris. During these tests, canisters were dried by heating externally by a heating blanket while simultaneously purging the canisters' interior with hot, dry nitrogen. Canister drying was found to be dominated by moisture retention properties of a concrete filler material (LICON) used for geometry control. This material extends the drying process 10 days or more beyond what would be required were it not there. The LICON resides in a nonpurgeable chamber separate from the core debris, and because of this configuration, dew point measurements on the exhaust stream do not provide a good indication of the dew point in the canisters. If the canisters are not dried, but rather just dewatered, 140-240 lb of water (not including the LICON water of hydration) will remain in each canister, approximately 50-110 lb of which is pore water in the LICON and the remainder unbound water

Shippingport Spent Fuel Canister System Description

International Nuclear Information System (INIS)

JOHNSON, D.M.

2000-01-01

In 1978 and 1979, a total of 72 blanket fuel assemblies (BFAs), irradiated during the operating cycles of the Shippingport Atomic Power Station's Pressurized Water Reactor (PWR) Core 2 from April 1965 to February 1974, were transferred to the Hanford Site and stored in underwater storage racks in Cell 2R at the 221-T Canyon (T-Plant). The initial objective was to recover the produced plutonium in the BFAs, but this never occurred and the fuel assemblies have remained within the water storage pool to the present time. The Shippingport Spent Fuel Canister (SSFC) is a confinement system that provides safe transport functions (in conjunction with the TN-WHC cask) and storage for the BFAs at the Canister Storage Building (CSB). The current plan is for these BFAs to be retrieved from wet storage and loaded into SSFCs for dry storage. The sealed SSFCs containing BFAs will be vacuum dried, internally backfilled with helium, and leak tested to provide suitable confinement for the BFAs during transport and storage. Following completion of the drying and inerting process, the SSFCs are to be delivered to the CSB for closure welding and long-term interim storage. The CSB will provide safe handling and dry storage for the SSFCs containing the BFAs. The purpose of this document is to describe the SSFC system and interface equipment, including the technical basis for the system, design descriptions, and operations requirements. It is intended that this document will be periodically updated as more equipment design and performance specification information becomes available

Defense High Level Waste Disposal Container System Description Document

International Nuclear Information System (INIS)

2000-01-01

The Defense High Level Waste 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 and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials

A review of the possible effects of hydrogen on lifetime of carbon steel nuclear waste canisters

International Nuclear Information System (INIS)

Turnbull, A.

2009-07-01

In Switzerland, the National Cooperative for the Disposal of Radioactive Waste (Nagra) is responsible for developing an effective method for the safe disposal of vitrified high level waste (HLW) and spent fuel. One of the options for disposal canisters is thick-walled carbon steel. The canisters, which would have a diameter of about 1 m and a length of about 3 m (HLW) or about 5 m (spent fuel), will be embedded in horizontal tunnels and surrounded with bentonite clay. The regulatory requirement for the minimum canister lifetime is 1000 years but demonstration of a minimum lifetime of 10,000 years would be desirable. The pore-water to which the canister will be exposed is of marine origin with about 0.1-0.3 M Cl-. Since hydrogen is generated during the corrosion process, it is necessary to assess the probability of hydrogen assisted cracking modes and to make recommendations to eliminate that probability. To that aim, key reports detailing projections for the local environment and associated corrosion rate of the waste canister have been evaluated with the focus on the implication for the absorbed hydrogen concentration in the steel. Simple calculations of hydrogen diffusion and accumulation in the inner compartment of the sealed canister indicate that a pressure equivalent to that for gas pockets external to the canister (envisaged to be about 10 MPa) may be attained in the proposed exposure time, an important consideration since it is not possible to modify the internal surface of the closure weld. Current ideas on mechanisms of hydrogen assisted cracking are assessed from which it is concluded that the mechanistic understanding and associated models of hydrogen assisted cracking are insufficient to provide a framework for quantitative prediction for this application. The emphasis then was to identify threshold conditions for cracking and to evaluate the likelihood that these may be exceeded over the lifetime of the containment. Based on an analysis of data in the

Analytical Evaluation of Preliminary Drop Tests Performed to Develop a Robust Design for the Standardized DOE Spent Nuclear Fuel Canister

International Nuclear Information System (INIS)

Ware, A.G.; Morton, D.K.; Smith, N.L.; Snow, S.D.; Rahl, T.E.

1999-01-01

The Department of Energy (DOE) has developed a design concept for a set of standard canisters for the handling, interim storage, transportation, and disposal in the national repository, of DOE spent nuclear fuel (SNF). The standardized DOE SNF canister has to be capable of handling virtually all of the DOE SNF in a variety of potential storage and transportation systems. It must also be acceptable to the repository, based on current and anticipated future requirements. This expected usage mandates a robust design. The canister design has four unique geometries, with lengths of approximately 10 feet or 15 feet, and an outside nominal diameter of 18 inches or 24 inches. The canister has been developed to withstand a drop from 30 feet onto a rigid (flat) surface, sustaining only minor damage - but no rupture - to the pressure (containment) boundary. The majority of the end drop-induced damage is confined to the skirt and lifting/stiffening ring components, which can be removed if de sired after an accidental drop. A canister, with its skirt and stiffening ring removed after an accidental drop, can continue to be used in service with appropriate operational steps being taken. Features of the design concept have been proven through drop testing and finite element analyses of smaller test specimens. Finite element analyses also validated the canister design for drops onto a rigid (flat) surface for a variety of canister orientations at impact, from vertical to 45 degrees off vertical. Actual 30-foot drop testing has also been performed to verify the final design, though limited to just two full-scale test canister drops. In each case, the analytical models accurately predicted the canister response

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.)

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

Unreviewed safety question evaluation of 100 K West fuel canister gas and liquid sampling

International Nuclear Information System (INIS)

Alwardt, L.D.

1995-01-01

The purpose of this report is to provide the basis for answers to an Unreviewed Safety Question (USQ) safety evaluation for the gas and liquid sampling activities associated with the fuel characterization program at the 100 K West (KW) fuel storage basin. The scope of this safety evaluation is limited to the movement of canisters between the main storage basin, weasel pit, and south loadout pit transfer channel (also known as the decapping station); gas and liquid sampling of fuel canisters in the weasel pit; mobile laboratory preliminary sample analysis in or near the 105 KW basin building; and the placement of sample containers in an approved shipping container. It was concluded that the activities and potential accident consequences associated with the gas and liquid sampling of 100 KW fuel canisters are bounded by the current safety basis documents and do not constitute an Unreviewed Safety Question

Inspection of bottom and lid welds for disposal canisters

International Nuclear Information System (INIS)

Pitkaenen, J.

2010-09-01

radiographic testing for this study. Other radiation sources cannot penetrate the wall thickness sufficiently. Defects detected in visual testing will be sized using eddy current testing. Special inspection techniques have been developed for sizing using low frequency (LF) eddy current testing. High frequency (HF) coils are used to evaluate the area of the surface defects. In the case of copper inspection, 30 kHz is considered a high frequency and 200 Hz a low frequency. The acceptance criteria define the defect detectability requirements of the inspection methods. The defect detectability of the methods is under further study to ascertain their viability. In the case of a single defect, the inaccuracy of the sizing method is less important. The significance of the inaccuracy in sizing is more pronounced when there are two or more defects in the corrosion path of the canister. The sizing capability will be studied and subsequently inspection techniques qualified on the basis of the recommendation of ENIQ (European Network for Inspection Qualification). This type of qualification is applied in Europe. In the USA, the qualification in nuclear applications is based on the performance demonstration part of ASME XI. All required information (input information, inspection procedures, technical justifications, etc.) will be gathered in order to prepare for the qualification. At the moment in the NDT reliability study the defect detectability is in focus as well the inaccuracy of sizing of indications. The precise sizing is the clearly more safety-based and industrial acceptance of canisters can be carried out. Sizing will be also verified by a limited metallographic study of detected indications. The work of technical justifications will be started. By qualifying the inspectors and inspection methods, the defined quality requirements for manufacturing disposal canisters for spent nuclear fuel will be met. (orig.)

Data compilation report: Gas and liquid samples from K West Basin fuel storage canisters

International Nuclear Information System (INIS)

Trimble, D.J.

1995-01-01

Forty-one gas and liquid samples were taken from spent fuel storage canisters in the K West Basin during a March 1995 sampling campaign. (Spent fuel from the N Reactor is stored in sealed canisters at the bottom of the K West Basin.) A description of the sampling process, gamma energy analysis data, and quantitative gas mass spectroscopy data are documented. This documentation does not include data analysis

Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements

International Nuclear Information System (INIS)

KLEM, M.J.

2000-01-01

In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8.0 References lists the

The prospective usage of the multi-purpose canister and impacts on the waste management and disposal system

International Nuclear Information System (INIS)

McLeod, N.B.

1993-01-01

The Multi-Purpose Canister (MPC) is designed to be loaded with spent fuel and sealed at reactors and then serve the functions of transport, storage and disposal without reopening. It can be either self-shielded or unshielded, thus requiring compatible overpacks for transport, storage and disposal. The MPC is not a new concept but it may now be viable because of the particular characteristics at Yucca Mountain: larger MPCs are possible because of ramp access to the repository horizon, and the less difficult temperature limits because of in-drift emplacement, rather than borehole emplacement. This paper describes the advantages and disadvantages of adopting the MPC as the principal technology to be employed in the US program. Use of the MPC permits integration of the utility and DOE portions of the system as well as among the elements within the DOE portion. Paradoxically, the principal disadvantage of the MPC is a direct consequence of its merit as an integrating technology. Full integration includes disposability without reopening, and requires that disposability design decisions be made and implemented well in advance of when waste package licensing uncertainties are resolved. There is, therefore, a risk that MPCs loaded prior to waste package licensing will have to be opened. This risk is discussed in terms of probability and consequences and various alternatives for mitigating this risk are discussed

2D and 3D thermal simulations for storage systems with internal natural convection for canistered spent fuel

International Nuclear Information System (INIS)

Yaksh, M.; Wang, C.

2004-01-01

In the US, the number of nuclear plants expected to implement on-site dry storage is increasing each year. As reactors burn advanced fuel assemblies to higher burnups, the dry storage systems will be required to accommodate higher heat loads. This is due to the increasing capacity of the systems and the need to store higher burnup fuel with reasonable cooling periods (i.e., five to six years). As the storage systems heat rejection design must be passive, natural convection is an efficient means for rejection of heat from the spent fuel to the surface of the canister boundary. The design presented in this paper is a canistered system that employs conduction, radiation and convection to reject heat from the canister, which is stored in a vertical concrete cask. The canister containing the spent fuel in this design is a right circular stainless steel vessel capable of storing 37 PWR fuel assemblies with a total canister heat load of 40 kW. Accompanying any design effort is the use of a numerical methodology that can accurately predict the peak-clad temperatures of the fuel and the structural components of the system. The main challenge to any analysis employing internal natural convection may be perceived as a practical limitation due to the size of the model. Since canisters are typically cylindrical, a two-dimensional model can be used to represent the canister. The fuel basket structure, which maintains the configuration of the spent fuel, is an array of square tubes, and is non-axisymmetric. Flow up through the fuel region in the basket encounters a complex cross section due to the fuel assembly rod array (up to 17 x 17). The flow region of the heated gas down the outside of the basket in the annulus between the canister shell and the basket assembly (downcomer) is also an irregular shaped area. To confirm that a two-dimensional (2D) modelling methodology is appropriate, a benchmark using results from a thermal test is required. The thermal test focuses on the

Physical properties of encapsulate spent fuel in canisters; Comportamiento fisico de las capsulas de almacenamiento

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-07-01

Spent fuel and high-level wastes will be permanently stored in a deep geological repository (AGP). Prior to this, they will be encapsulated in canisters. The present report is dedicated to the study of such canisters under the different physical demands that they may undergo, be those in operating or accident conditions. The physical demands of interest include mechanical demands, both static and dynamic, and thermal demands. Consideration is given to the complete file of the canister, from the time when it is empty and without lid to the final conditions expected in the repository. Thermal analyses of canisters containing spent fuel are often carried out in two dimensions, some times with hypotheses of axial symmetry and some times using a plane transverse section through the centre of the canister. The results obtained in both types of analyses are compared here to those of complete three-dimensional analyses. The latter generate more reliable information about the temperatures that may be experienced by the canister and its contents; they also allow calibrating the errors embodied in the two-dimensional calculations. (Author)

Near-field performance of the advanced cold process canister

International Nuclear Information System (INIS)

Werme, L.

1991-12-01

A near-field performance evaluation of an advanced cold process canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie local-iced corrosion for the steel or copper canisters can be dismissed as a failure mechanism; The evaluation of the effects of processed outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. For completeness also evaluation of post-failure behaviour was carried out. Analyses were focussed on low probability phenomena from faults in canisters. Some items were identified where further research is justified in order to increase knowledge of the phenomena and thus strengthen the confidence of safety margins. However, it can be concluded that the risks of these scenarios can be judged to be acceptable. This is due to the fact that firstly, the probability of occurrence of most of these scenarios can be controlled to a large extent through technical measures. Secondly, these analyses indicated that the consequences would not be severe

Treatment and final disposal of nuclear waste. Programme for encapsulation, deep geological disposal, and research, development and demonstration

International Nuclear Information System (INIS)

1995-09-01

Programs for RD and D concerning disposal of radioactive waste are presented. Main topics include: Design, testing and manufacture of canisters for the spent fuels; Design of equipment for deposition of waste canisters; Material and process for backfilling rock caverns; Evaluation of accuracy and validation of methods for safety analyses; Development of methods for defining scenarios for the safety analyses. 471 refs, 67 figs, 21 tabs

Treatment and final disposal of nuclear waste. Programme for encapsulation, deep geological disposal, and research, development and demonstration

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

Programs for RD and D concerning disposal of radioactive waste are presented. Main topics include: Design, testing and manufacture of canisters for the spent fuels; Design of equipment for deposition of waste canisters; Material and process for backfilling rock caverns; Evaluation of accuracy and validation of methods for safety analyses; Development of methods for defining scenarios for the safety analyses. 471 refs, 67 figs, 21 tabs.

Stress redistribution and void growth in butt-welded canisters for spent nuclear fuel

International Nuclear Information System (INIS)

Josefson, B.L.; Karlsson, L.; Haeggblad, H.Aa.

1993-02-01

The stress-redistribution in Cu-Fe canisters for spent nuclear fuel during waiting for deposition and after final deposition is calculated numerically. The constitutive equation modelling creep deformation during this time period employs values on materials parameters determined within the SKB-project on 'mechanical integrity of canisters for spent nuclear fuel'. The welding residual stresses are redistributed without lowering maximum values during the waiting period, a very low amount of void growth is predicted for this type of copper during the deposition period. This leads to an estimated very large rupture time

Oxidative dissolution of spent fuel and release of nuclides from a copper/iron canister. Model developments and applications

Energy Technology Data Exchange (ETDEWEB)

Longcheng Liu

2001-12-01

Three models have been developed and applied in the performance assessment of a final repository. They are based on accepted theories and experimental results for known and possible mechanisms that may dominate in the oxidative dissolution of spent fuel and the release of nuclides from a canister. Assuming that the canister is breached at an early stage after disposal, the three models describe three sub-systems in the near field of the repository, in which the governing processes and mechanisms are quite different. In the model for the oxidative dissolution of the fuel matrix, a set of kinetic descriptions is provided that describes the oxidative dissolution of the fuel matrix and the release of the embedded nuclides. In particular, the effect of autocatalytic reduction of hexavalent uranium by dissolved H{sub 2}, using UO{sub 2} (s) on the fuel pellets as a catalyst, is taken into account. The simulation results suggest that most of the radiolytic oxidants will be consumed by the oxidation of the fuel matrix, and that much less will be depleted by dissolved ferrous iron. Most of the radiolytically produced hexavalent uranium will be reduced by the autocatalytic reaction with H{sub 2} on the fuel surface. It will reprecipitate as UO{sub 2} (s) on the fuel surface, and thus very little net oxidation of the fuel will take place. In the reactive transport model, the interactions of multiple processes within a defective canister are described, in which numerous redox reactions take place as multiple species diffuse. The effect of corrosion of the cast iron insert of the canister and the reduction of dissolved hexavalent uranium by ferrous iron sorbed onto iron corrosion products and by dissolved H{sub 2} are particularly included. Scoping calculations suggest that corrosion of the iron insert will occur primarily under anaerobic conditions. The escaping oxidants from the fuel rods will migrate toward the iron insert. Much of these oxidants will, however, be consumed

Hanford Spent Nuclear Fuel Project evaluation of multi-canister overpack venting and monitoring options during staging of K basins fuel

Energy Technology Data Exchange (ETDEWEB)

Wiborg, J.C.

1995-12-01

This engineering study recommends whether multi-canister overpacks containing spent nuclear fuel from the Hanford K Basins should be staged in vented or a sealed, but ventable, condition during staging at the Canister Storage Building prior to hot vacuum conditioning and interim storage. The integrally related issues of MCO monitoring, end point criteria, and assessing the practicality of avoiding venting and Hot Vacuum Conditioning for a portion of the spent fuel are also considered.

K West Basin canister survey

International Nuclear Information System (INIS)

Pitner, A.L.

1998-01-01

A survey was conducted of the K West Basin to determine the distribution of canister types that contain the irradiated N Reactor fuel. An underwater camera was used to conduct the survey during June 1998, and the results were recorded on videotape. A full row-by-row survey of the entire basin was performed, with the distinction between aluminum and stainless steel Mark 1 canisters made by the presence or absence of steel rings on the canister trunions (aluminum canisters have the steel rings). The results of the survey are presented in tables and figures. Grid maps of the three bays show the canister lid ID number and the canister type in each location that contained fuel. The following abbreviations are used in the grid maps for canister type designation: IA = Mark 1 aluminum, IS = Mark 1 stainless steel, and 2 = Mark 2 stainless steel. An overall summary of the canister distribution survey is presented in Table 1. The total number of canisters found to contain fuel was 3842, with 20% being Mark 1 Al, 25% being Mark 1 SS, and 55% being Mark 2 SS. The aluminum canisters were predominantly located in the East and West bays of the basin

Development and demonstration of prototype transportation equipment for emplacing HL vitrified waste canisters into small diameter bored horizontal disposal cells

International Nuclear Information System (INIS)

Seidler, Wolf K.; Bosgiraud, Jean-Michel; Londe, Louis

2008-01-01

Over a period of 4 and years the National Radioactive Waste Management Agency (Andra), working with a variety of Contractors mostly specializing in nuclear orientated mechanical applications, successfully designed, fabricated and demonstrated 2 very different prototype high level waste transport systems. The first system, based on air cushion technology, was developed primarily for very heavy loads (17 to 45 tonnes). The results of this work are described in a separate presentation (Paper 21) at this Conference. The second system, developed by Andra within the framework of the ESDRED Project, generally referred to as the 'Pushing Robot System' for vitrified waste canisters, is the subject of this paper. The 'Pushing Robot System' is a part of the French national disposal concept that is described in Andra's 'Dossier 2005'. The latter is a public document that can be viewed on Andra's web site (www.andra.fr). The 'Pushing Robot System' system is designed for the deep geological disposal (in clay formations) of 'C' type vitrified waste canisters. In its entirety the system provides for the transport, emplacement and, if necessary, the retrieval of those canisters. Nothing in the design of the Andra emplacement equipment would preclude its utilization in horizontal openings in other types of geological settings. Over a period of some 8 years Andra has developed the 'Pushing Robot System' in 3 phases. Initially there was only the 'Conceptual Design' (Phase 1) which was incorporated in the Dossier 2005. This was followed by Phase 2 i.e. the design and fabrication of a simplified full scale prototype system henceforth referred to a P1, which includes a Pushing Robot, a Dummy Canister and a Test Bench. P1 details were also incorporated in the Dossier 2005. Finally, during Phase 3, a second more comprehensive full scale prototype system P2 has been designed and is being assembled and tested this month. This system includes a Transport Shuttle, a Transfer Shielding Cask, a

Comparison of Tagging Technologies for Safeguards of Copper Canisters for Nuclear Spent Fuel.

Science.gov (United States)

Clementi, Chiara; Littmann, François; Capineri, Lorenzo

2018-03-21

Several countries are planning to store nuclear spent fuel in long term geological repositories, preserved by copper canisters with an iron insert. This new approach involves many challenging problems and one is to satisfy safeguards requirements: the Continuity of Knowledge (CoK) of the fuel must be kept from the encapsulation plant up to the final repository. To date, no measurement system has been suggested for a unique identification and authentication. Following the list of the most important safeguards, safety and security requirements for copper canisters identification and authentication, a review of conventional tagging technologies and measurement systems for nuclear items is reported in this paper. The aim of this study is to verify to what extent each technology could be potentially used for keeping the CoK of copper canisters. Several tagging methods are briefly described and compared, discussing advantages and disadvantages.

Spent nuclear fuel Canister Storage Building CDR Review Committee report

International Nuclear Information System (INIS)

Dana, W.P.

1995-12-01

The Canister Storage Building (CSB) is a subproject under the Spent Nuclear Fuels Major System Acquisition. This subproject is necessary to design and construct a facility capable of providing dry storage of repackaged spent fuels received from K Basins. The CSB project completed a Conceptual Design Report (CDR) implementing current project requirements. A Design Review Committee was established to review the CDR. This document is the final report summarizing that review

Corrosion of canister materials for radioactive waste disposal

International Nuclear Information System (INIS)

Kienzler, Bernhard

2017-01-01

In the period between 1980 and 2004, corrosion studies on various metallic materials have been performed at the Research Center Karlsruhe. The objectives of these experimental studies addressed mainly the performance of canister materials for heat producing, high-level wastes and spent nuclear fuels for a repository in a German salt dome. Additional studies covered the performance of steels for packaging wastes with negligible heat production under conditions to be expected in rocksalt and in the Konrad iron ore mine. The results of the investigations have been published in journals and conference proceedings but also in ''grey literature''. This paper presents a summary of the results of corrosion experiments with fine-grained steels and nodular cast steel.

HLW Disposal System Development

Energy Technology Data Exchange (ETDEWEB)

Choi, J. W.; Choi, H. J.; Lee, J. Y. (and others)

2007-06-15

A KRS is suggested through design requirement analysis of the buffer and the canister which are the constituent of disposal system engineered barrier and HLW management plans are proposed. In the aspect of radionuclide retention capacity, the thickness of the buffer is determined 0.5m, the shape to be disc and ring and the dry density to be 1.6 g/cm{sup 3}. The maximum temperature of the buffer is below 100 .deg. which meets the design requirement. And bentonite blocks with 5 wt% of graphite showed more than 1.0 W/mK of thermal conductivity without the addition of sand. The result of the thermal analysis for proposed double-layered buffer shows that decrease of 7 .deg. C in maximum temperature of the buffer. For the disposal canister, the copper for the outer shell material and cast iron for the inner structure material is recommended considering the results analyzed in terms of performance of the canisters and manufacturability and the geochemical properties of deep groundwater sampled from the research area with granite, salt water intrusion, and the heavy weight of the canister. The results of safety analysis for the canister shows that the criticality for the normal case including uncertainty is the value of 0.816 which meets subcritical condition. Considering nation's 'Basic Plan for Electric Power Demand and Supply' and based on the scenario of disposing CANDU spent fuels in the first phase, the disposal system that the repository will be excavated in eight phases with the construction of the Underground Research Laboratory (URL) beginning in 2020 and commissioning in 2040 until the closure of the repository is proposed. Since there is close correlation between domestic HLW management plans and front-end/back-end fuel cycle plans causing such a great sensitivity of international environment factor, items related to assuring the non-proliferation and observing the international standard are showed to be the influential factor and acceptability

Vacuum Drying Tests for Storage of Aluminum Spent Nuclear Fuel

International Nuclear Information System (INIS)

Chen, K.F.; Large, W.S.; Sindelar, R.L.

1998-05-01

A total inventory of up to approximately 32,000 aluminum-based spent nuclear fuel (Al SNF) assemblies are expected to be shipped to Savannah River Site (SRS) from domestic and foreign research reactors over the next several decades. Treatment technologies are being developed as alternatives to processing for the ultimate disposition of Al SNF in the geologic repository. One technology, called Direct/Co-disposal of Al SNF, would place the SNF into a canister ready for disposal in a waste package, with or without canisters containing high-level radioactive waste glass logs, in the repository. The Al SNF would be transferred from wet storage and would need to be dried in the Al SNF canister. The moisture content inside the Al SNF canister is limited to avoid excessive Al SNF corrosion and hydrogen buildup during interim storage before disposal. A vacuum drying process was proposed to dry the Al SNF in a canister. There are two major concerns for the vacuum drying process. One is water inside the canister could become frozen during the vacuum drying process and the other one is the detection of dryness inside the canister. To vacuum dry an irradiated fuel in a heavily shielded canister, it would be very difficult to open the lid to inspect the dryness during the vacuum drying operation. A vacuum drying test program using a mock SNF assembly was conducted to demonstrate feasibility of drying the Al SNF in a canister. These tests also served as a check-out of the drying apparatus for future tests in which irradiated fuel would be loaded into a canister under water followed by drying for storage

Spent nuclear fuel project multi-canister overpack, additional NRC requirements

International Nuclear Information System (INIS)

Garvin, L.J.

1998-01-01

The US Department of Energy (DOE), established in the K Basin Spent Nuclear Fuel Project Regulatory Policy, dated August 4, 1995 (hereafter referred to as the Policy), the requirement for new Spent Nuclear Fuel (SNF) Project facilities to achieve nuclear safety equivalency to comparable US Nuclear Regulatory Commission (NRC)-licensed facilities. For activities other than during transport, when the Multi-Canister Overpack (MCO) is used and resides in the Canister Storage Building (CSB), Cold Vacuum Drying (CVD) facility or Hot Conditioning System, additional NRC requirements will also apply to the MCO based on the safety functions it performs and its interfaces with the SNF Project facilities. An evaluation was performed in consideration of the MCO safety functions to identify any additional NRC requirements needed, in combination with the existing and applicable DOE requirements, to establish nuclear safety equivalency for the MCO. The background, basic safety issues and general comparison of NRC and DOE requirements for the SNF Project are presented in WHC-SD-SNF-DB-002

Design of double containment canister cask storage system

International Nuclear Information System (INIS)

Asami, M.; Matsumoto, T.; Oohama, T.; Kuriyama, K.; Kawakami, K.

2004-01-01

Spent fuels discharged from Japanese LWR will be stored as recycled-fuel-resources in interim storage facilities. The concrete cask storage system is one of important forms for the spent fuel interim storage. In Japan, the interim storage facility will be located near the coast, therefore it is important to prevent SCC (Stress Corrosion Cracking) caused by sea salt particles and to assure the containment integrity of the canister which contains spent fuels. KEPCO, NFT and OCL have designed the double containment canister cask storage system that can assure the long-term containment integrity and monitor the containment performance without storage capacity decrease. Major features of the combined canister cask system are shown as follows: This system can survey containment integrity of dual canisters by monitoring the pressure of the gap between canisters. The primary canister has dual lids sealed by welding. The secondary canister has single lid tightened by bolts and sealed by metallic gaskets. The primary canister is contained in the transport cask during transportation, and the gap between the primary canister and the transport cask is filled with He gas. Under storage condition in the concrete cask, the primary canister is contained in the secondary canister, and the gap between these canisters is filled with helium gas. Hence this system can prevent the primary canister to contact sea salt particle in the air and from SCC. Decrease of cooling performance because of the double canister is compensated by fins fitted on the secondary canister surface. Then, this system can prevent the decrease of storage capacity determined by the fuel temperature limit. This system can assure that the primary canister will keep intact for long term storage. Therefore, in the case of pressure down of the gap between canisters, it can be considered that the secondary canister containment is damaged, and the primary canister will be transferred to another secondary canister at the

Development of a Korean Reference disposal System(A-KRS) for the HLW from Advanced Fuel Cycles

International Nuclear Information System (INIS)

Choi, Heui Joo; Choi, J. W.; Lee, J. Y.

2010-04-01

A database program for analyzing the characteristics of spent fuels was developed, and A-SOURCE program for characterizing the source term of HLW from advanced fuel cycles. A new technique for developing a copper canister by introducing a cold spray technique was developed, which could reduce the amount of copper. Also, to enhance the performance of A-KRS, two kinds of properties, thermal performance and iodine adsorption, were studied successfully. A complex geological disposal system which can accommodate all the HLW (CANDU and HANARO spent fuels, HLW from pyro-processing of PWR spent fuels, decommissioning wastes) was developed, and a conceptual design was carried out. Operational safety assessment system was constructed for the long-term management of A-KRS. Three representative accidental cases were analyzed, and the probabilistic safety assessment was adopted as a methodology for the safety evaluation of A-KRS operation. A national program was proposed to support the HLW national policy on the HLW management. A roadmap for HLW management was proposed based on the optimum timing of disposal

Facility Description 2012. Summary report of the encapsulation plant and disposal facility designs

International Nuclear Information System (INIS)

Palomaeki, J.; Ristimaeki, L.

2013-10-01

The purpose of the facility description is to be a specific summary report of the scope of Posiva's nuclear facilities (encapsulation plant and disposal facility) in Olkiluoto. This facility description is based on the 2012 designs and completing Posiva working reports. The facility description depicts the nuclear facilities and their operation as the disposal of spent nuclear fuel starts in Olkiluoto in about 2020. According to the decisions-in-principle of the government, the spent nuclear fuel from Loviisa and Olkiluoto nuclear power plants in operation and in future cumulative spent nuclear fuel from Loviisa 1 and 2, Olkiluoto 1, 2, 3 and 4 nuclear power plants, is permitted to be disposed of in Olkiluoto bedrock. The design of the disposal facility is based on the KBS-3V concept (vertical disposal). Long-term safety concept is based on the multi-barrier principle i.e. several release barriers, which ensure one another so that insufficiency in the performance of one barrier doesn't jeopardize long-term safety of the disposal. The release barriers are the following: canister, bentonite buffer and deposition tunnel backfill, and the host rock around the repository. The canisters are installed into the deposition holes, which are bored to the floor of the deposition tunnels. The canisters are enveloped with compacted bentonite blocks, which swell after absorbing water. The surrounding bedrock and the central and access tunnel backfill provide additional retardation, retention, and dilution. The nuclear facilities consist of an encapsulation plant and of underground final disposal facility including other aboveground buildings and surface structures serving the facility. The access tunnel and ventilation shafts to the underground disposal facility and some auxiliary rooms are constructed as a part of ONKALO underground rock characterization facility during years 2004-2014. The construction works needed for the repository start after obtaining the construction

Corrosion resistance of a copper canister for spent nuclear fuel

International Nuclear Information System (INIS)

1983-04-01

The report presents an evaluation of copper as canister material for spent nuclear fuel. The evaluation is made from the viewpoint of corrosion and applies to a concept of 1977. Supplementary corrosion studies have been performed. The report includes 9 appendices which deal with experimental data. (G.B.)

Design of the disposal facility 2012

International Nuclear Information System (INIS)

Saanio, T.; Ikonen, A.; Keto, P.; Kirkkomaeki, T.; Kukkola, T.; Nieminen, J.; Raiko, H.

2013-11-01

The spent nuclear fuel accumulated from the nuclear power plants in Olkiluoto in Eurajoki and in Haestholmen in Loviisa will be disposed of in Olkiluoto. A facility complex will be constructed at Olkiluoto, and it will include two nuclear waste facilities according to Government Degree 736/2008. The nuclear waste facilities are an encapsulation plant, constructed to encapsulate spent nuclear fuel and a disposal facility consisting of an underground repository and other underground rooms and above ground service spaces. The repository is planned to be excavated to a depth of 400 - 450 meters. Access routes to the disposal facility are an inclined access tunnel and vertical shafts. The encapsulated fuel is transferred to the disposal facility in the canister lift. The canisters are transferred from the technical rooms to the disposal area via central tunnel and deposited in the deposition holes which are bored in the floors of the deposition tunnels and are lined beforehand with compacted bentonite blocks. Two parallel central tunnels connect all the deposition tunnels and these central tunnels are inter-connected at regular intervals. The solution improves the fire safety of the underground rooms and allows flexible backfilling and closing of the deposition tunnels in stages during the operational phase of the repository. An underground rock characterization facility, ONKALO, is excavated at the disposal level. ONKALO is designed and constructed so that it can later serve as part of the repository. The goal is that the first part of the disposal facility will be constructed under the building permit phase in the 2010's and operations will start in the 2020's. The fuel from 4 operating reactors as well the fuel from the fifth nuclear power plant under construction, has been taken into account in designing the disposal facility. According to the information from TVO and Fortum, the amount of the spent nuclear fuel is 5,440 tU. The disposal facility is being excavated

Corrosion of canister materials for radioactive waste disposal

Energy Technology Data Exchange (ETDEWEB)

Kienzler, Bernhard [KIT Karlsruhe (Germany). Institut fuer Nukleare Entsorgung (INE)

2017-08-15

In the period between 1980 and 2004, corrosion studies on various metallic materials have been performed at the Research Center Karlsruhe. The objectives of these experimental studies addressed mainly the performance of canister materials for heat producing, high-level wastes and spent nuclear fuels for a repository in a German salt dome. Additional studies covered the performance of steels for packaging wastes with negligible heat production under conditions to be expected in rocksalt and in the Konrad iron ore mine. The results of the investigations have been published in journals and conference proceedings but also in ''grey literature''. This paper presents a summary of the results of corrosion experiments with fine-grained steels and nodular cast steel.

Hazard categorization of 100 K West fuel canister gas and liquid sampling

International Nuclear Information System (INIS)

Alwardt, L.D.

1994-01-01

This report documents the determination that the activities associated with the 100 K West fuel canister gas and liquid sampling are classified as Hazard Category Other (consequences are below criteria for Category 3)

Stakeholder involvement in the evaluation of a multipurpose canister system

International Nuclear Information System (INIS)

Williams, J.R.; Kane, D.; Smith, T.B. Jr.

1994-01-01

The U.S. Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM), began evaluating a multipurpose canister (MPC) concept in October of 1992. This followed recommendations by the Nuclear Waste Technical Review Board (NWTRB) and the U.S. Nuclear Regulatory Commission (NRC) that DOE develop a nuclear waste management system that achieves system integration, standardization, and reduced fuel-handling operations. Industry organizations such as Edison Electric Institute (EEI) and Electric Power Research Institute (EPRI) had conducted earlier studies that concluded advantages to the nuclear waste management system may be offered by such a concept. The MPC concept involves a metal canister which would contain multiple spent nuclear fuel assemblies. The canister would be sealed at the nuclear power plant and would not be reopened. The MPC would then be placed inside separate casks or overpacks for storage, transportation, and disposal. An important factor in DOE's evaluation of the MPC concept was the involvement of external parties. This paper describes that involvement process for the OCRWM's MPC implementation program. External parties who have an interest or stake in the program are referred to as stakeholders

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

DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

International Nuclear Information System (INIS)

Radulesscu, G.; Tang, J.S.

2000-01-01

The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M andO [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M andO 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M andQ 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M andO 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable canisters. The intended use of this

DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

Energy Technology Data Exchange (ETDEWEB)

G. Radulesscu; J.S. Tang

2000-06-07

The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable

Technical note. A review of the mechanical integrity of the canister

International Nuclear Information System (INIS)

Segle, Peter

2012-01-01

Background: The Swedish Radiation Safety Authority (SSM) reviews the Swedish Nuclear Fuel Company's (SKB) applications under the Act on Nuclear Activities (SFS 1984:3) for the construction and operation of a repository for spent nuclear fuel and for an encapsulation facility. As part of the review, SSM commissions consultants to carry out work in order to obtain information on specific issues. The results from the consultants' tasks are reported in SSM's Technical Note series. Objectives of the project: This project is part of SSM:s review of SKB:s license application for final disposal of spent nuclear fuel. The assignment concerns a review of the mechanical integrity of the canister. Summary by the author: An introductory review of SR-Site has been conducted with respect to the mechanical integrity of the canister. The review is focused on the copper canister and the nodular cast iron insert. Review results show that a number of loads and loading scenarios for the copper canister has not been analysed by SKB. The importance of sufficient creep ductility of the copper material and sufficient ductility and fracture toughness of the nodular cast iron material is pointed out in the review. A sensitivity study is suggested where the impact of these properties on the mechanical integrity of the canister is investigated. It is also suggested that potential damage mechanisms influencing these properties are further investigated. SKB's modelling of creep elongation at rupture under repository conditions is questioned. Needs for complementary information from SKB for the main review of SR-Site is listed. A list of review topics for SSM is also suggested

Test manufacture of a canister insert

International Nuclear Information System (INIS)

Raiko, H.

2004-11-01

This report describes the insert-manufacturing test of a disposal canister for spent nuclear fuel that was made by Metso Paper Oy, Jyvaeskylae Foundry, in 2003 on contract for Posiva Oy. The test manufacture was a part of the co-operation development programme of encapsulation technology between SKB AB and Posiva Oy. Insert casting was specified according to the current manufacturing specifications of SKB. The canister insert was of BWR-type with integral bottom. This was the first trial manufacture of this type of insert in Finland and, in total, the second test manufacture of insert by Metso Paper. The result fulfilled all the requirements but the material mechanical properties and metallurgical structure of the cast material. The measured tensile strength, ultimate strength and elongation at rupture were lower than specified. The reason for this was revealed in the metallurgical investigation of the cast material. The nodulizing of the graphite was not occurred during the casting process according to the requirements. (orig.)

Mechanical Integrity of Copper Canister Lid and Cylinder. Sensitivity study

International Nuclear Information System (INIS)

Karlsson, Marianne

2002-08-01

This report is part of a study of the mechanical integrity of canisters used for disposal of nuclear fuel waste. The overall objective is to determine and ensure the static and long-term strength of the copper canister lid and cylinder casing. The canisters used for disposal nuclear fuel waste of type BWR consists of an inner part (insert) of ductile cast iron and an outer part of copper. The copper canister is to provide a sealed barrier between the contents of the canister and the surroundings. The study in this report complements the finite element analyses performed in an earlier study. The analyses aim to evaluate the sensitivity of the canister to tolerances regarding the gap between the copper cylinder and the cast iron insert. Since great uncertainties regarding the material's long term creep properties prevail, analyses are also performed to evaluate the effect of different creep data on the resulting strain and stress state. The report analyses the mechanical response of the lid and flange of the copper canister when subjected to loads caused by pressure from swelling bentonite and from groundwater at a depth of 500 meter. The loads acting on the canister are somewhat uncertain and the cases investigated in this report are possible cases. Load cases analysed are: Pressure 15 MPa uniformly distributed on lid and 5 MPa uniformly distributed on cylinder; Pressure 5 MPa uniformly distributed on lid and 15 MPa uniformly distributed on cylinder; Pressure 20 MPa uniformly distributed on lid and cylinder; and Side pressures 10 MPa and 20 MPa uniformly distributed on part of the cylinder. Creep analyses are performed for two of the load cases. For all considered designs high principal stresses appear on the outside of the copper cylinder in the region from the weld down to the level of the lid lower edge. Altering the gap between lid and cylinder and/or between cylinder and insert only marginally affects the resulting stress state. Fitting the lid in the cylinder

Mechanical Integrity of Canisters Using a Fracture Mechanics Approach

Energy Technology Data Exchange (ETDEWEB)

Koyama, Tomofumi; Guoxiang Zhang; Lanru Jing [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Land and Water Resources Engineering

2006-07-15

This report presents the methods and results of a research project about numerical modeling of mechanical integrity of cast-iron canisters for the final disposal of spent nuclear fuel in Sweden, using combined boundary element (BEM) and finite element (FEM) methods. The objectives of the project are: 1) to investigate the possibility of initiation and growth of fractures in the cast-iron canisters under the mechanical loading conditions defined in the premises of canister design by Swedish Nuclear Fuel and Waste Management Co. (SKB); 2) to investigate the maximum bearing capacity of the cast iron canisters under uniformly distributed and gradually increasing boundary pressure until plastic failure. Achievement of the two objectives may provide some quantitative evidence for the mechanical integrity and overall safety of the cast-iron canisters that are needed for the final safety assessment of the geological repository of the radioactive waste repository in Sweden. The geometrical dimension, distribution and magnitudes of loads and Material properties of the canisters and possible fractures were provided by the latest investigations of SKB. The results of the BEM simulations, using the commercial code BEASY, indicate that under the currently defined loading conditions the possibility of initiation of new fractures or growth of existing fractures (defects) are very small, due to the reasons that: 1) the canisters are under mainly compressive stresses; 2) the induced tensile stress regions are too small in both dimension and magnitude to create new fractures or to induce growth of existing fractures, besides the fact that the toughness of the fractures in the cast iron canisters are much higher that the stress intensity factors in the fracture tips. The results of the FEM simulation show a approximately 75 MPa maximum pressure beyond which plastic collapse of the cast-iron canisters may occur, using an elastoplastic Material model. This figure is smaller compared

Transport of multiassembly sealed canisters

International Nuclear Information System (INIS)

Quinn, R.D.; Lehnert, R.A.; Rosa, J.M.

1992-01-01

A significant portion of the commercial spent nuclear fuel in dry storage in the US will be stored in multiassembly sealed canisters before the DOE begins accepting fuel from utilities in 1998. This paper reports that it is desirable from economic and ALARA perspectives to transfer these canisters directly from the plant to the MRS. To this end, it is necessary that the multiassembly sealed canisters, which have been licensed for storage under 10CFR72, be qualified for shipment within a suitable shipping cask under the rules of 10CFR71. Preliminary work performed to date indicates that it is feasible to license a current canister design for transportation, and work is proceeding on obtaining NRC approval

Probabilistic analysis and material characterisation of canister insert for spent nuclear fuel. Summary report

Energy Technology Data Exchange (ETDEWEB)

Andersson, Claes-Goeran [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Andersson, Mats; Erixon, Bo [AaF Industriteknik, Stockholm (Sweden); Bjoerkegren, Lars-Erik [Swedish Foundry Association, Stockholm (Sweden); Dillstroem, Peter [DNV Technology, Stockholm (Sweden); Minnebo, Philip

2005-11-15

The KBS-3 canister for geological disposal of spent nuclear fuel in Sweden consists of a ductile cast iron insert and a copper shielding. The canister should inhibit release of radionuclides for at least 100,000 years. The copper protects the canister from corrosion whereas the ductile cast iron insert provides the mechanical strength. In the repository the hydrostatic pressure from the groundwater and the swelling pressure from the surrounding bentonite, which in total results in a maximum pressure of 14 MPa, will load the canisters in compression. During the extreme time scales, ice ages are expected with a maximum ice thickness of 3,000 m resulting in an additional pressure of 30 MPa. The maximum design pressure for the KBS-3 canisters has therefore been set to be 44 MPa. A relatively large number of canisters have been manufactured as part of SKB's development programme. To verify the strength of the canisters at this stage of development SKB initiated a project in cooperation with the European commissions Joint Research Centre (JRC), Institute of Energy in Petten in the Netherlands, together with a number of other partners. Three inserts manufactured by different Swedish foundries were used in the project. A large statistical test programme was developed to determine statistical distributions of various material parameters and defect distributions. These data together with the results from stress and strain finite element analysis were subsequently used in probabilistic analysis to determine the probability for plastic collapse caused by high pressure or fracture by crack growth in regions with tensile stresses. The main conclusions from the probabilistic analysis are: 1. At the design pressure of 44 MPa, the probability of failure is insignificant ({approx}2x10{sup -9}). This is the case even though several conservative assumptions have been made. 2. The stresses in the insert caused by the outer pressure are mainly compressive. The regions with tensile

Effect of Canister Movement on Water Turbidity

International Nuclear Information System (INIS)

TRIMBLE, D.J.

2000-01-01

Requirements for evaluating the adherence characteristics of sludge on the fuel stored in the K East Basin and the effect of canister movement on basin water turbidity are documented in Briggs (1996). The results of the sludge adherence testing have been documented (Bergmann 1996). This report documents the results of the canister movement tests. The purpose of the canister movement tests was to characterize water turbidity under controlled canister movements (Briggs 1996). The tests were designed to evaluate methods for minimizing the plumes and controlling water turbidity during fuel movements leading to multi-canister overpack (MCO) loading. It was expected that the test data would provide qualitative visual information for use in the design of the fuel retrieval and water treatment systems. Video recordings of the tests were to be the only information collected

Research on corrosion aspects of the advanced cold process canister

International Nuclear Information System (INIS)

Blackwood, D.J.; Hoch, A.R.; Naish, C.C.; Rance, A.

1994-01-01

The Advanced Cold Process Canister (ACPC) is a waste canister being developed jointly by SKB and TVO for the disposal of spent nuclear fuel. It comprises an outer copper canister, with a carbon steel canister inside. A concern regarding the use of the ACPC is that, in the unlikely event that the outer copper canister is penetrated, the anaerobic corrosion of the carbon steel container may result in the formation of hydrogen gas bubbles. These bubbles could disrupt the backfill, and thus increase water flow through the near field and the flux of radionuclides to the host geology. A number of factors that influence the rate at which hydrogen evolves as a result of the anaerobic corrosion of carbon steel in artificial granitic groundwaters have been investigated. A previously observed, time-dependent decline in the hydrogen evolution rate has been confirmed as being due to the production of magnetite film. Once the magnetite film is about 0.7-1.0 μm thick, the rate of hydrogen evolution reaches a steady state value. The pH and the ionic strength of the groundwater were both found to influence the long-term hydrogen evolution rate. The results of the experimental programme were used to update a model of the corrosion behaviour and hydrogen production from the Advanced Cold Process Canister. 36 figs, 5 tabs, 13 refs

Examination of sludge from the Hanford K Basins fuel canisters

International Nuclear Information System (INIS)

Makenas, B.J.

1998-01-01

Samples of sludges with a high uranium content have been retrieved from the fuel canisters in the Hanford K West and K East basins. The composition of these samples contrasts markedly with the previously reported content of sludge samples taken from the K East basin floor. Chemical composition, chemical reactivity, and particle size of sludge are summarized in this paper

Deep Borehole Disposal as an Alternative Concept to Deep Geological Disposal

Energy Technology Data Exchange (ETDEWEB)

Lee, Jongyoul; Lee, Minsoo; Choi, Heuijoo; Kim, Kyungsu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this paper, the general concept and key technologies for deep borehole disposal of spent fuels or HLW, as an alternative method to the mined geological disposal method, were reviewed. After then an analysis on the distance between boreholes for the disposal of HLW was carried out. Based on the results, a disposal area were calculated approximately and compared with that of mined geological disposal. These results will be used as an input for the analyses of applicability for DBD in Korea. The disposal safety of this system has been demonstrated with underground research laboratory and some advanced countries such as Finland and Sweden are implementing their disposal project on commercial stage. However, if the spent fuels or the high-level radioactive wastes can be disposed of in the depth of 3-5 km and more stable rock formation, it has several advantages. Therefore, as an alternative disposal concept to the mined deep geological disposal concept (DGD), very deep borehole disposal (DBD) technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, the general concept of deep borehole disposal for spent fuels or high level radioactive wastes was reviewed. And the key technologies, such as drilling technology of large diameter borehole, packaging and emplacement technology, sealing technology and performance/safety analyses technologies, and their challenges in development of deep borehole disposal system were analyzed. Also, very preliminary deep borehole disposal concept including disposal canister concept was developed according to the nuclear environment in Korea.

Deep Borehole Disposal as an Alternative Concept to Deep Geological Disposal

International Nuclear Information System (INIS)

Lee, Jongyoul; Lee, Minsoo; Choi, Heuijoo; Kim, Kyungsu

2016-01-01

In this paper, the general concept and key technologies for deep borehole disposal of spent fuels or HLW, as an alternative method to the mined geological disposal method, were reviewed. After then an analysis on the distance between boreholes for the disposal of HLW was carried out. Based on the results, a disposal area were calculated approximately and compared with that of mined geological disposal. These results will be used as an input for the analyses of applicability for DBD in Korea. The disposal safety of this system has been demonstrated with underground research laboratory and some advanced countries such as Finland and Sweden are implementing their disposal project on commercial stage. However, if the spent fuels or the high-level radioactive wastes can be disposed of in the depth of 3-5 km and more stable rock formation, it has several advantages. Therefore, as an alternative disposal concept to the mined deep geological disposal concept (DGD), very deep borehole disposal (DBD) technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, the general concept of deep borehole disposal for spent fuels or high level radioactive wastes was reviewed. And the key technologies, such as drilling technology of large diameter borehole, packaging and emplacement technology, sealing technology and performance/safety analyses technologies, and their challenges in development of deep borehole disposal system were analyzed. Also, very preliminary deep borehole disposal concept including disposal canister concept was developed according to the nuclear environment in Korea

Engineered Barrier System - Mechanical Integrity of KBS-3 Spent Fuel Canisters. Report from a Workshop. Synthesis and extended abstracts

Energy Technology Data Exchange (ETDEWEB)

2007-09-15

SKI is preparing to review the license applications being developed by the Swedish Nuclear Fuel and Waste Management Company (SKB) for a final repository for the geological disposal of spent nuclear fuel in the year 2009. As part of its preparation, SKI is conducting a series of technical workshops on key aspects of the Engineered Barrier System (EBS). The workshop reported here mainly dealt with the mechanical integrity of KBS-3 spent fuel canisters. This included assessment and review of various loading conditions, structural integrity models and mechanical properties of the copper shell and the cast iron insert. Degradation mechanisms such as stress corrosion cracking and brittle creep fracture were also briefly addressed. Previous workshops have addressed the overall concept for long-term integrity of the EBS, the manufacturing, testing and QA of the EBS, the performance confirmation for the EBS, long-term stability of the buffer and the backfill, corrosion properties of copper canisters and the spent fuel dissolution and source term modelling. The goal of ongoing review work in connection of the workshop series is to achieve a comprehensive overview of all aspects of SKB's EBS and spent fuel work prior to the handling of the forthcoming license application. This report aims to summarise the issues discussed at the workshop and to extract the essential viewpoints that have been expressed. The report is not a comprehensive record of all the discussions at the workshop, and individual statements made by workshop participants should be regarded as personal opinions rather than SKI viewpoints. Results from the EBS workshops series will be used as one important basis in future review work. This reports includes in addition to the workshop synthesis, questions to SKB identified prior to the workshop, and extended abstracts for introductory presentations

The disposal of Canada's nuclear fuel waste: engineering for a disposal facility

International Nuclear Information System (INIS)

Simmons, G.R.; Baumgartner, P.

1994-01-01

This report presents some general considerations for engineering a nuclear fuel waste disposal facility, alternative disposal-vault concepts and arrangements, and a conceptual design of a used-fuel disposal centre that was used to assess the technical feasibility, costs and potential effects of disposal. The general considerations and alternative disposal-vault arrangements are presented to show that options are available to allow the design to be adapted to actual site conditions. The conceptual design for a used-fuel disposal centre includes descriptions of the two major components of the disposal facility, the Used-Fuel Packaging Plant and the disposal vault; the ancillary facilities and services needed to carry out the operations are also identified. The development of the disposal facility, its operation, its decommissioning, and the reclamation of the site are discussed. The costs, labour requirements and schedules used to assess socioeconomic effects and that may be used to assess the cost burden of waste disposal to the consumer of nuclear energy are estimated. The Canadian Nuclear Fuel Waste Management Program is funded jointly by AECL and Ontario Hydro under the auspices of the CANDU Owners Group. (author)

SRL canister impact tests

International Nuclear Information System (INIS)

Kelker, J.W. Jr.

1986-05-01

The Defense Waste Processing Facility (DWPF) is being constructed at the SRP for the containerization of high-level nuclear waste as a waste form for eventual permanent disposal. The waste will be incorporated in molten glass and solidified in Type 304L stainless steel canisters 2 feet in diameter x 9 feet 10 inches long. The canisters have a minimum wall thickness of 3/8 inch. Over a three-year period, nineteen drop-tests of nine canisters, filled with simulated waste glass, were made in support of the DWPF containerization program. Eight of the canister evaluation tests were of Type 304L stainless steel material and one was of commercially pure titanium. Three different length (9.44, 5.06, and 7.88 inch) nozzle configurations containing final closure upset welds were evaluated for the stainless steel canisters. All impact tests of the stainless steel canisters, which included bottom-, side-, and top-drops, were acceptable. The bottom-drop test of the titanium canister, which contained a final closure upset weld, was acceptable; however, the top-drop resulted in a breaching of the top head where it joins the nozzle. The final closure titanium upset weld was acceptable. The titanium canister wall thickness was 1/4 inch

Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

International Nuclear Information System (INIS)

PICKETT, W.W.

2000-01-01

The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure

Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

Energy Technology Data Exchange (ETDEWEB)

PICKETT, W.W.

2000-09-22

The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure.

An assessment of KW Basin radionuclide activity when opening SNF canisters

International Nuclear Information System (INIS)

Bergmann, D.W.; Mollerus, F.J.; Wray, J.L.

1995-01-01

N Reactor spent fuel is being stored in sealed canisters in the KW Basin. Some of the canisters contain damaged fuel elements. There is the potential for release of Cs 137, Kr 85, H3, and other fission products and transuranics (TRUs) when canisters are opened. Canister opening is required to select and transfer fuel elements to the 300 Area for examination as part of the Spent Nuclear Fuel (SNF) Characterization program. This report estimates the amount of radionuclides that can be released from Mark II spent nuclear fuel (SNF) canisters in KW Basin when canisters are opened for SNF fuel sampling as part of the SNF Characterization Program. The report also assesses the dose consequences of the releases and steps that can be taken to reduce the impacts of these releases

Feasibility of using a high-level waste canister as an engineered barrier in disposal

International Nuclear Information System (INIS)

Slate, S.C.; Pitman, S.G.; Nesbitt, J.F.; Partain, W.L.

1982-08-01

The objective of this report is to evaluate the feasibility of designing a process canister that could also serve as a barrier canister. To do this a general set of performance criteria is assumed and several metal alloys having a high probability of demonstrating high corrosion resistance under repository conditions are evaluated in a qualitative design assessment. This assessment encompasses canister manufacture, the glass-filling process, interim storage, transportation, and to a limited extent, disposal in a repository. A series of scoping tests were carried out on two titanium alloys and Inconel 625 to determine if the high temperature inherent in the glass-fill processing would seriously affect either the strength or corrosion resistance of these metals. This is a process-related concern unique to the barrier canister concept. The material properties were affected by the heat treatments which simulated both the joule-heated glass melter process (titanium alloys and Inconel 625) and the in-can melter (ICM) process (Inconel 625). However, changes in the material properties were generally within 20% of the original specimens. Accelerated corrosion testing of the heat treated coupons in a highly oxygenated brine showed basic corrosion resistance of titanium grade 12 and Inconel 625 to compare favorably with that of the untreated coupons. The titanium grade 2 coupons experienced severe corrosion pitting. These corrosion tests were of a scoping nature and suitable primarily for the detection of gross sensitivity to the heat treatment inherent in the glass-fill process. They are only suggstive of repository performance since the tests do not adequately model the wide range of repository conditions that could conceivably occur

A reference container concept for spent fuel disposal: structural safety for dimensioning of the reference container

International Nuclear Information System (INIS)

Choi, Jongwon; Kwon, Sangki; Kang, Chulhyung; Kwon, Youngjoo

2002-01-01

This paper presents the mechanical and thermal stress analysis of a disposal canister to provide basic information for dimensioning the canister and configuration of the canister components. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head, and the thermal load build up in the container

High-level waste processing and disposal

International Nuclear Information System (INIS)

Crandall, J.L.; Krause, H.; Sombret, C.; Uematsu, K.

1984-11-01

Without reprocessing, spent LWR fuel itself is generally considered an acceptable waste form. With reprocessing, borosilicate glass canisters, have now gained general acceptance for waste immobilization. The current first choice for disposal is emplacement in an engineered structure in a mined cavern at a depth of 500-1000 meters. A variety of rock types are being investigated including basalt, clay, granite, salt, shale, and volcanic tuff. This paper gives specific coverage to the national high level waste disposal plans for France, the Federal Republic of Germany, Japan and the United States. The French nuclear program assumes prompt reprocessing of its spent fuels, and France has already constructed the AVM. Two larger borosilicate glass plants are planned for a new French reprocessing plant at La Hague. France plans to hold the glass canisters in near-surface storage for a forty to sixty year cooling period and then to place them into a mined repository. The FRG and Japan also plan reprocessing for their LWR fuels. Both are currently having some fuel reprocessed by France, but both are also planning reprocessing plants which will include waste vitrification facilities. West Germany is now constructing the PAMELA Plant at Mol, Belgium to vitrify high level reprocessing wastes at the shutdown Eurochemic Plant. Japan is now operating a vitrification mockup test facility and plans a pilot plant facility at the Tokai reprocessing plant by 1990. Both countries have active geologic repository programs. The United State program assumes little LWR fuel reprocessing and is thus primarily aimed at direct disposal of spent fuel into mined repositories. However, the US have two borosilicate glass plants under construction to vitrify existing reprocessing wastes

Welding of the lid and the bottom of the disposal canister

International Nuclear Information System (INIS)

Meuronen, I.; Salonen, T.

2010-10-01

The seal welding of the lid and bottom of a copper disposal canister for spent nuclear fuel using ordinary electron beam welding (EBW) made in a vacuum and the results gained in the development work are presented in this report. As an alternative method, the friction stir welding (FSW) is also presented in an overview. Welding of copper is very challenging mainly due to the high thermal conductivity of the copper material. The EBW method is based on so-called deep penetration welding which does not use additional welding material. The convenience of the method is that the weld is the same material as the base material. When compared to other fusion welding methods, the material transitions in the material caused by EBW are slight. The EBW process typically has a high number of welding parameters but, in practice, only a few parameters are adjusted during copper welding to maintain weld quality and the stability of the process. The high vacuum required by the method prevents the material from oxidising but, on the other hand, it narrows the application of the method. The requirements presented for the weld and welding process can be divided in two classes. The first class contains the requirements intended to ensure the long-term safety of the canister. Corrosion resistance and adequate creep ductility are such requirements. The second class requirements correspond to welding process requirements for component manufacture, the components themselves and the other processes of the encapsulation plant. The welding process, including the personnel, equipment and process validation, shall also fulfil the special requirements concerning all nuclear plants in general. The quality assurance and control (QA/QC) for welding is presented as a separate section. The welding quality assurance contains the personnel, equipment and the welding process. For EBW process validation there are available norms and acceptation procedures. In these, the essential component is the

SITE-94. CAMEO: A model of mass-transport limited general corrosion of copper canisters

International Nuclear Information System (INIS)

Worgan, K.J.; Apted, M.J.

1996-12-01

This report describes the technical basis for the CAMEO code, which models the general, uniform corrosion of a copper canister either by transport of corrodants to the canister, or by transport of corrosion products away from the canister. According to the current Swedish concept for final disposal of spent nuclear fuels, extremely long containment times are achieved by thick (60-100 mm) copper canisters. Each canister is surrounded by a compacted bentonite buffer, located in a saturated, crystalline rock at a depth of around 500 m below ground level. Three diffusive transport-limited cases are identified for general, uniform corrosion of copper: General corrosion rate-limited by diffusive mass-transport of sulphide to the canister surface under reducing conditions; General corrosion rate-limited by diffusive mass-transport of oxygen to the canister surface under mildly oxidizing conditions; General corrosion rate-limited by diffusive mass-transport of copper chloride away from the canister surface under highly oxidizing conditions. The CAMEO code includes general corrosion models for each of the above three processes. CAMEO is based on the well-tested CALIBRE code previously developed as a finite-difference, mass-transfer analysis code for the SKI to evaluate long-term radionuclide release and transport in the near-field. A series of scoping calculations for the general, uniform corrosion of a reference copper canister are presented

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

International Nuclear Information System (INIS)

1990-03-01

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

Fuel and canister process report for the safety assessment SR-Site

International Nuclear Information System (INIS)

Werme, Lars; Lilja, Christina

2010-12-01

This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

Fuel and canister process report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

Werme, Lars; Lilja, Christina (eds.)

2010-12-15

This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

Evaluation of the heat transfer in a geological repository concept containing PWR, VHTR and hybrid ads-fission spent fuels

Energy Technology Data Exchange (ETDEWEB)

Jonusan, Raoni A.S.; Pereira, Fernando; Velasquez, Carlos E.; Salome, Jean A.D.; Cardoso, Fabiano; Pereira, Claubia; Fortini, Angela, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-11-01

The investigation of the thermal behavior of spent fuel (SF) materials is essential to determining appropriate potential sites to accommodate geological repositories as well as the design of canisters, considering their potential risk to people health and of environmental contamination. This work presents studies of the temperature in a canister containing spent fuels discharged from Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) reactor systems in a geological repository concept. The thermal analyses were performed with the software ANSYS, which is widely used to solve engineering problems through the Finite Element Method. The ANSYS Transient Thermal module was used. The spent nuclear fuels were set as heat sources using data of previous studies derived from decay heat curves. The studies were based on comparison of the mean temperature on a canister surface along the time under geological disposal conditions, for a same amount of each type of spent nuclear fuel evaluated. The results conclude that fuels from VHTR and ADS systems are inappropriate to be disposed in a standardized PWR canister, demanding new studies to determine the optimal amount of spent fuel and new internal canister geometries. It is also possible to conclude that the hypothetical situation of a single type of canister being used to accommodate different types of spent nuclear fuels is not technically feasible. (author)

EB-welding of the copper canister for the nuclear waste disposal. Final report of the development programme 1994-1997

International Nuclear Information System (INIS)

Aalto, H.

1998-10-01

During 1994-1997 Posiva Oy and Outokumpu Poricopper Oy had a joint project Development of EB-welding method for massive copper canister manufacturing. The project was part of the national technology program 'Weld 2000' and it was supported financially by Technology Development Centre (TEKES). The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper canisters and placed deep into the bedrock. The thick copper layer of the canister provides a long time corrosion resistance and prevents deposited nuclear fuel from contact with water. The quality requirements of the copper components are high because of the designed long lifetime of the canister. The EB-welding technology has proved to be applicable method for the production of the copper canisters and the EB-welding technique is needed at least when the lids of the copper canister will be closed. There are a number of parameters in EB-welding which affect weldability. However, the effect of the welding parameters and their optimization has not been extensively studied in welding of thick copper sections using conventional high vacuum EB-welding. One aim of this development work was to extensively study effect of welding parameters on weld quality. The final objective was to minimise welding defects in the main weld and optimize slope out procedure in thick copper EB-welding. Welding of 50 mm thick copper sections was optimized using vertical and horizontal EB-welding techniques. As a result two full scale copper lids were welded to a short cylinder successfully. The resulting weld quality with optimised welding parameters was reasonable good. The optimised welding parameters for horizontal and vertical beam can be applied to the longitudinal body welds of the canister. The optimal slope out procedure for the lid closure needs some additional development work. In addition of extensive EB-welding program ultrasonic inspection and creep strength of the weld were studied. According

EB-welding of the copper canister for the nuclear waste disposal. Final report of the development programme 1994-1997

Energy Technology Data Exchange (ETDEWEB)

Aalto, H. [Outokumpu Oy Poricopper, Pori (Finland)

1998-10-01

During 1994-1997 Posiva Oy and Outokumpu Poricopper Oy had a joint project Development of EB-welding method for massive copper canister manufacturing. The project was part of the national technology program `Weld 2000` and it was supported financially by Technology Development Centre (TEKES). The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper canisters and placed deep into the bedrock. The thick copper layer of the canister provides a long time corrosion resistance and prevents deposited nuclear fuel from contact with water. The quality requirements of the copper components are high because of the designed long lifetime of the canister. The EB-welding technology has proved to be applicable method for the production of the copper canisters and the EB-welding technique is needed at least when the lids of the copper canister will be closed. There are a number of parameters in EB-welding which affect weldability. However, the effect of the welding parameters and their optimization has not been extensively studied in welding of thick copper sections using conventional high vacuum EB-welding. One aim of this development work was to extensively study effect of welding parameters on weld quality. The final objective was to minimise welding defects in the main weld and optimize slope out procedure in thick copper EB-welding. Welding of 50 mm thick copper sections was optimized using vertical and horizontal EB-welding techniques. As a result two full scale copper lids were welded to a short cylinder successfully. The resulting weld quality with optimised welding parameters was reasonable good. The optimised welding parameters for horizontal and vertical beam can be applied to the longitudinal body welds of the canister. The optimal slope out procedure for the lid closure needs some additional development work. In addition of extensive EB-welding program ultrasonic inspection and creep strength of the weld were studied. According

Safeguards for final disposal of spent nuclear fuel. Methods and technologies for the Olkiluoto site

International Nuclear Information System (INIS)

Okko, O.

2003-05-01

The final disposal of the nuclear material shall introduce new safeguards concerns which have not been addressed previously in IAEA safeguards approaches for spent fuel. The encapsulation plant to be built at the site will be the final opportunity for verification of spent fuel assemblies prior to their transfer to the geological repository. Moreover, additional safety and safeguards measures are considered for the underground repository. Integrated safeguards verification systems will also concentrate on environmental monitoring to observe unannounced activities related to possible diversion schemes at the repository site. The final disposal of spent nuclear fuel in geological formation will begin in Finland within 10 years. After the geological site investigations and according to legal decision made in 2001, the final repository of the spent nuclear fuel shall be located at the Olkiluoto site in Eurajoki. The next phase of site investigations contains the construction of an underground facility, called ONKALO, for rock characterisation purposes. The excavation of the ONKALO is scheduled to start in 2004. Later on, the ONKALO may form a part of the final repository. The plans to construct the underground facility for nuclear material signify that the first safeguards measures, e.g. baseline mapping of the site area, need to take prior to the excavation phase. In order to support the development and implementation of the regulatory control of the final disposal programme, STUK established an independent expert group, LOSKA. The group should support the STUK in the development of the technical safeguards requirements, in the implementation of the safeguards and in the evaluation of the plans of the facility operator. This publication includes four background reports produced by this group. The first of these 'NDA verification of spent fuel, monitoring of disposal canisters, interaction of the safeguards and safety issues in the final disposal' describes the new

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

SPENT NUCLEAR FUEL NUMBER DENSITIES FOR MULTI-PURPOSE CANISTER CRITICALITY CALCULATIONS

International Nuclear Information System (INIS)

D. A. Thomas

1996-01-01

The purpose of this analysis is to calculate the number densities for spent nuclear fuel (SNF) to be used in criticality evaluations of the Multi-Purpose Canister (MPC) waste packages. The objective of this analysis is to provide material number density information which will be referenced by future MPC criticality design analyses, such as for those supporting the Conceptual Design Report

NDE to Manage Atmospheric SCC in Canisters for Dry Storage of Spent Fuel: An Assessment

Energy Technology Data Exchange (ETDEWEB)

Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pardini, Allan F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cuta, Judith M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Adkins, Harold E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qiao, Hong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Larche, Michael R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Diaz, Aaron A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Doctor, Steven R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2013-09-01

This report documents efforts to assess representative horizontal (Transuclear NUHOMS®) and vertical (Holtec HI-STORM) storage systems for the implementation of non-destructive examination (NDE) methods or techniques to manage atmospheric stress corrosion cracking (SCC) in canisters for dry storage of used nuclear fuel. The assessment is conducted by assessing accessibility and deployment, environmental compatibility, and applicability of NDE methods. A recommendation of this assessment is to focus on bulk ultrasonic and eddy current techniques for direct canister monitoring of atmospheric SCC. This assessment also highlights canister regions that may be most vulnerable to atmospheric SCC to guide the use of bulk ultrasonic and eddy current examinations. An assessment of accessibility also identifies canister regions that are easiest and more difficult to access through the ventilation paths of the concrete shielding modules. A conceivable sampling strategy for canister inspections is to sample only the easiest to access portions of vulnerable regions. There are aspects to performing an NDE inspection of dry canister storage system (DCSS) canisters for atmospheric SCC that have not been addressed in previous performance studies. These aspects provide the basis for recommendations of future efforts to determine the capability and performance of eddy current and bulk ultrasonic examinations for atmospheric SCC in DCSS canisters. Finally, other important areas of investigation are identified including the development of instrumented surveillance specimens to identify when conditions are conducive for atmospheric SCC, characterization of atmospheric SCC morphology, and an assessment of air flow patterns over canister surfaces and their influence on chloride deposition.

Status of the Multipurpose Canister (MPC) Project

International Nuclear Information System (INIS)

Hopper, J.P.

1996-01-01

The multipurpose canister (MPC) project represents a cornerstone of the current U.S. Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) program for handling spent nuclear fuel. The MPC and associated support equipment is being designed to accommodate the requirements for not only storage and transport but also for the specified disposal requirements of the mined geologic repository system. The phase 1 design effort for the MPC system, being performed by the Westinghouse team on behalf of TRW Environmental Safety Systems (TESS), the OCRWM management ampersand operating (M ampersand O) contractor, is on schedule for delivery of completed safety analysis reports (SARs) in April 1996

Multi-purpose canister system evaluation: A systems engineering approach

International Nuclear Information System (INIS)

1994-09-01

This report summarizes Department of Energy (DOE) efforts to investigate various container systems for handling, transporting, storing, and disposing of spent nuclear fuel (SNF) assemblies in the Civilian Radioactive Waste Management System (CRWMS). The primary goal of DOE's investigations was to select a container technology that could handle the vast majority of commercial SNF at a reasonable cost, while ensuring the safety of the public and protecting the environment. Several alternative cask and canister concepts were evaluated for SNF assembly packaging to determine the most suitable concept. Of these alternatives, the multi-purpose canister (MPC) system was determined to be the most suitable. Based on the results of these evaluations, the decision was made to proceed with design and certification of the MPC system. A decision to fabricate and deploy MPCs will be made after further studies and preparation of an environmental impact statement

Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

Energy Technology Data Exchange (ETDEWEB)

BAZINET, G.D.

2001-05-15

The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those requirements are noted in section 3

Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

Energy Technology Data Exchange (ETDEWEB)

BAZINET, G.D.

2000-11-03

The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. The purpose of this revision is to document completion of verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those

Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

International Nuclear Information System (INIS)

BAZINET, G.D.

2001-01-01

The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those requirements are noted in section 3.1.5 and will be

Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

International Nuclear Information System (INIS)

BAZINET, G.D.

2000-01-01

The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. The purpose of this revision is to document completion of verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those requirements are noted

Chemical and mineralogical aspects of water-bentonite interaction in nuclear fuel disposal conditions

International Nuclear Information System (INIS)

Melamed, A.; Pitkaenen, P.

1996-01-01

In the field of nuclear fuel disposal, bentonite has been selected as the principal sealing and buffer material for placement around waste canisters, forming both a mechanical and chemical barrier between the radioactive waste and the surrounding ground water. Ion exchange and mineral alteration processes were investigated in a laboratory study of the long-term interaction between compacted Na-bentonite (Volclay MX-80) and ground water solutions, conducted under simulated nuclear fuel disposal conditions. The possible alteration of montmorillonite into illite has been a major object of the mineralogical study. However, no analytical evidence was found, that would indicate the formation of this non-expandable clay type. Apparently, the change of montmorillonite from Na- to Ca-rich was found to be the major alteration process in bentonite. In the water, a concentration decrease in Ca, Mg, and K, and an increase in Na, HCO 3 and SO 4 were recorded. The amount of calcium ions available in the water was considered insufficient to account for the recorded formation of Ca-montmorillonite. It is therefore assumed that the accessory Ca-bearing minerals in bentonite provide the fundamental source of these cations, which exchange with sodium during the alteration process. (38 refs.)

Test manufacture of the canister insert 135

International Nuclear Information System (INIS)

Raiko, H.

2005-10-01

This report describes the insert-manufacturing test of a disposal canister for spent nuclear fuel that was made by Metso Foundries Jyvaeskylae Oy, in June 2004 on contract for Posiva Oy. The test manufacture was a part of the co-operation development programme of encapsulation technology between SKB AB and Posiva Oy. Insert casting was specified according to the current manufacturing specifications of SKB. The canister insert was of BWR-type with integral bottom. This was the second trial manufacture of this type of insert in Finland and, in total, the third test manufacture of insert by Metso Foundries Jyvaeskylae Oy. The result fulfilled all the requirements but the material mechanical properties of the cast material. The measured ultimate strength and elongation at rupture were lower than specified in the upper part of the cast. The reason for this was revealed in the metallurgical investigation of the cast material. The cast contained slag (dross). Avoiding the dross formation will be the most demanding challenge of the forthcoming development of the cast procedure. (orig.)

Spent nuclear fuel project product specification

International Nuclear Information System (INIS)

Pajunen, A.L.

1998-01-01

Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification

Waste disposal package

Science.gov (United States)

Smith, M.J.

1985-06-19

This is a claim for a waste disposal package including an inner or primary canister for containing hazardous and/or radioactive wastes. The primary canister is encapsulated by an outer or secondary barrier formed of a porous ceramic material to control ingress of water to the canister and the release rate of wastes upon breach on the canister. 4 figs.

Canister materials proposed for final disposal of high level nuclear waste

Energy Technology Data Exchange (ETDEWEB)

Mattson, E; Odoj, R; Merz, E [eds.

1981-06-01

The nuclear waste will be enclosed in corrosion resistant canisters. These will be deposited in repositories in geological formations, such as granite, basalt, clay, bedded or domed salt, or the sediments beneath the deep ocean floor. There the canisters will be exposed to groundwater, brine or seawater at an elevated temperature. Species formed by radiolysis may effect the corrosivity of the agent. The corrosion resistance of candidate canister materials is evaluated by corrosion tests and by thermodynamic and mass transport calculations. Examinations of ancient metal objects after long exposure in nature may give additional information. On the basis of the work carried out so far, the principal candidate canister materials are titanium materials, copper, and highpurity alumina.

Corrosion studies on HGW-canister materials for marine disposal

International Nuclear Information System (INIS)

Taylor, K.J.; Bland, I.D.; Marsh, G.P.

1984-07-01

A combination of mathematical modelling and experimental studies has been used to investigate and assess the long term corrosion behaviour of heat generating waste canister/ overpack materials under conditions relevant to deep ocean disposal. Preliminary operation of the model, using improved electrochemical kinetic data from the experimental programme, has indicated that the general corrosion rate of carbon steel at 90 deg C will be 57 μm yr -1 which is equivalent to a metal loss of 57 mm in 1000 years. This prediction compares favourably with the results from long term tests, which are also in progress, for plain and electron beam welded carbon steel specimens embedded in marine sediment at 90 deg C under active dissolution conditions. Tests with γ-radiation at a dose rate of 1.5 x 10 5 R h -1 have shown that the pH of seawater falls to 3.7 after 5000 hours exposure causing a significant increase in the corrosion rate of carbon steel from 50 to 80 μm yr -1 . Further work is in progress to investigate the mechanism of this acidification and whether it also occurs at the more realistic lower radiation dose rates. (author)

Research of radioactive waste storage cask/canister materials, spent nuclear fuels and various radioactive waste forms and development of their assessment methods. Final report for Stage 3

International Nuclear Information System (INIS)

Dobrev, D.; Balek, V.; Červinka, R.; Večerník, P.; Člupek, M.; Kouřil, M.; Novák, P.; Stoulil, J.; Silber, R.

2013-08-01

The main topics treated are: Research and development of methodologies for canister/cask material degradation assessment; Laboratory research of selected materials of canister/cask with radioactive waste; and Research and assessment of canister/cask materials in natural granite rocks. Two additional documents are appended: Corrosion rate determination for samples in compacted bentonite in anaerobic conditions (methodology), and Roll test for corrosion test in an occluded solution at the interface between a radioactive waste disposal canister and the bentonite cover. (P.A.)

A Preliminary Assessment of a Deep Borehole disposal of Spent Fuels

International Nuclear Information System (INIS)

Lee, Younmyoung; Jeon, Jongtae

2014-01-01

Deep borehole disposal (DBD) of such radioactive waste as spent nuclear fuels (SFs) and other waste forms has been investigating mainly at Sandia National Labs for the US DOE as an alternative option. DBD can give advantages over less deep geological disposal since the disposal of wastes at a great depth where a low degree of permeability in the potentially steady rock condition will be beneficial for nuclide movement. Groundwater in the deep basement rock can even have salinity and less chance to mix with groundwater above. The DBD concept is quite straightforward and even simple: Waste canisters are simply emplaced in the lower 2 km part of the borehole down to 5 km deep. Through this study, a conceptual DBD is assessed for a similar case as the US DOE's approach, in which 400 SF canisters are to be emplaced at a deep bottom between 3km and 5km depths, upon which an additional 1km-thick compacted bentonite is overbuffered, and the remaining upper part of the borehole is backfilled again with a mixture of crushed rock and bentonite. Then, the total 5km-deep borehole has three zones: a disposal zone at the bottom 2km, a buffer zone at the next 1km, and backfill zone at the rest top 2km, as illustrated conceptually in Fig. 1. To demonstrate the feasibility in view of long-term radiological safety, a rough model for a safety assessment of this conceptual deep borehole repository system, providing detailed models for nuclide transport in and around the geosphere and biosphere under normal nuclide release scenarios that can occur after a closure of the repository, has been developed using GoldSim. A simple preliminary result in terms of the dose exposure rate from a safety assessment of the DBD is also presented and compared to the case of direct disposal of SFs in a KBS-3V vertical type repository, carried out in previous studies. For different types and shapes of repositories at each different depth, direct comparison between a DBD and a KBS-3 type disposal of

A Preliminary Assessment of a Deep Borehole disposal of Spent Fuels

Energy Technology Data Exchange (ETDEWEB)

Lee, Younmyoung; Jeon, Jongtae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Deep borehole disposal (DBD) of such radioactive waste as spent nuclear fuels (SFs) and other waste forms has been investigating mainly at Sandia National Labs for the US DOE as an alternative option. DBD can give advantages over less deep geological disposal since the disposal of wastes at a great depth where a low degree of permeability in the potentially steady rock condition will be beneficial for nuclide movement. Groundwater in the deep basement rock can even have salinity and less chance to mix with groundwater above. The DBD concept is quite straightforward and even simple: Waste canisters are simply emplaced in the lower 2 km part of the borehole down to 5 km deep. Through this study, a conceptual DBD is assessed for a similar case as the US DOE's approach, in which 400 SF canisters are to be emplaced at a deep bottom between 3km and 5km depths, upon which an additional 1km-thick compacted bentonite is overbuffered, and the remaining upper part of the borehole is backfilled again with a mixture of crushed rock and bentonite. Then, the total 5km-deep borehole has three zones: a disposal zone at the bottom 2km, a buffer zone at the next 1km, and backfill zone at the rest top 2km, as illustrated conceptually in Fig. 1. To demonstrate the feasibility in view of long-term radiological safety, a rough model for a safety assessment of this conceptual deep borehole repository system, providing detailed models for nuclide transport in and around the geosphere and biosphere under normal nuclide release scenarios that can occur after a closure of the repository, has been developed using GoldSim. A simple preliminary result in terms of the dose exposure rate from a safety assessment of the DBD is also presented and compared to the case of direct disposal of SFs in a KBS-3V vertical type repository, carried out in previous studies. For different types and shapes of repositories at each different depth, direct comparison between a DBD and a KBS-3 type disposal of

Analysis for Eccentric Multi Canister Overpack (MCO) Drops at the Canister Storage Building (CSB) (CSB-S-0073)

Energy Technology Data Exchange (ETDEWEB)

TU, K.C.

1999-10-08

Multi-Canister Overpacks (MCOs) containing spent nuclear fuel (SNF) will be routinely handled at the Canister Storage Building (CSB) during fuel movement operations in the SNF Project. This analysis was performed to investigate the potential for damage from an eccentric accidental drop onto the standard storage tube, overpack tube, service station, or sample/weld station. Appendix D was added to the FDNW document to include the peer Review Comment Record & transmittal record.

Analysis for Eccentric Multi Canister Overpack (MCO) Drops at the Canister Storage Building (CSB) (CSB-S-0073)

International Nuclear Information System (INIS)

TU, K.C.

1999-01-01

Multi-Canister Overpacks (MCOs) containing spent nuclear fuel (SNF) will be routinely handled at the Canister Storage Building (CSB) during fuel movement operations in the SNF Project. This analysis was performed to investigate the potential for damage from an eccentric accidental drop onto the standard storage tube, overpack tube, service station, or sample/weld station. Appendix D was added to the FDNW document to include the peer Review Comment Record and transmittal record

FPIN2 posttest analysis of cylindrical canisters in SLSF Experiment P4

Energy Technology Data Exchange (ETDEWEB)

Hughes, T H; Kramer, J M

1984-12-01

Results demonstrate that the clad deformation is dominated by the expansion of the fuel when it melts. In our analysis we moved the end space volume and some of the fuel-clad radial gap volume to an artificial central hole. This approximation may affect the details in the early parts of the transient, but clearly did not affect the major cladding deformation. It is also clear that the accuracy of the value of the fuel expansion upon melting is significant as is the dimensional accuracy of the fuel and canisters. The major conclusions from the FPIN2 posttest analysis of the cylindrical canisters in SLSF Experiment P4 are: The maximum melt fractions in the two canisters were about 75%. Both canisters experienced about the same diametral strains of 12% prior to failure. These strains were almost entirely due to the additional volume that must be created inside the canisters to accommodate the expansion of fuel on melting. The mode of cladding failure was plastic instability by necking of the canister walls. The failure time of the 20% CW canister and the nonmechanical failure of the 10% CW canister are consistent with the FPIN2 calculations using the plastic instability failure criteria.

Waste package designs for disposal of high-level waste in salt formations

International Nuclear Information System (INIS)

Basham, S.J. Jr.; Carr, J.A.

1984-01-01

In the United States of America the selected method for disposal of radioactive waste is mined repositories located in suitable geohydrological settings. Currently four types of host rocks are under consideration: tuff, basalt, crystalline rock and salt. Development of waste package designs for incorporation in mined salt repositories is discussed. The three pertinent high-level waste forms are: spent fuel, as disassembled and close-packed fuel pins in a mild steel canister; commercial high-level waste (CHLW), as borosilicate glass in stainless-steel canisters; defence high-level waste (DHLW), as borosilicate glass in stainless-steel canisters. The canisters are production and handling items only. They have no planned long-term isolation function. Each waste form requires a different approach in package design. However, the general geometry and the materials of the three designs are identical. The selected waste package design is an overpack of low carbon steel with a welded closure. This container surrounds the waste forms. Studies to better define brine quantity and composition, radiation effects on the salt and brines, long-term corrosion behaviour of the low carbon steel, and the leaching behaviour of the spent fuel and borosilicate glass waste forms are continuing. (author)

Canister transfer into repository in shaft alternative

International Nuclear Information System (INIS)

Raiko, H.; Kukkola, T.; Autio, J.

2005-09-01

In this report, a study of lift transportation of a massive canister for spent nuclear fuel is considered. The canister is transferred from ground level to repository, which lies in the depth of 400 to 500 m in the bedrock. The canister is a massive metal vessel, whose weight is 19 to 29 tons, and which is strongly irradiant (gamma and neutrons), and which contains 1.4 to 2.2 tons of very strongly radio-active material, the activity of the fuel should not be spread in the environment even during postulated accidents. The study observes that the lift alternative is possible to be built and through good design practices and good maintenance procedures its safety, reliability and usability can be kept on such high level that canister transport is estimated to be licensable. (orig.)

Method of ground disposal of radioactive waste

International Nuclear Information System (INIS)

Harashina, Heihachi.

1991-01-01

Rock bases are drilled to form a disposal hole, an overhanging hole and a burying hole each as a shaft. An appropriate number of canisters prepared by vitrification of high level radioactive wastes are charged in the disposal hole with a gap to the inner wall of the hole. Shock absorbers each made of bentonite are filled between each of the canisters and between the canister and the inner wall of the disposal hole, and the canisters are entirely covered with the layer of the shock absorbers. Further, plucking materials having water sealing property such as cement mortar are filled thereover. With such a constitution, in a case if water should intrude into the overhung portion, since the disposal hole is covered with the large flange portion in addition to the water sealing performance of the plucking, the shock absorbers and the canisters undergo no undesirable effects. Further, in a case if water should intrude to the disposal hole, the shock absorber layers are swollen by water absorption, to suppress the intrusion of water. (T.M.)

Reduction of uranium in disposal conditions of spent nuclear fuel

International Nuclear Information System (INIS)

Myllykylae, E.

2008-02-01

This literature study is a summary of publications, in which the reduction of uranium by iron has been investigated in anaerobic groundwater conditions or in aqueous solution in general. The basics of the reduction phenomena and the oxidation states, complexes and solubilities of uranium and iron in groundwaters are discussed as an introduction to the subject, as well as, the Finnish disposal concept of spent nuclear fuel. The spent fuel itself mainly (∼96 %) consists of a sparingly soluble uranium(IV) dioxide, UO 2 (s), which is stable phase in the anticipated reducing disposal conditions. If spent fuel gets in contact with groundwater, oxidizing conditions might be induced by the radiolysis of water, or by the intrusion of oxidizing glacial melting water. Under these conditions, the oxidation and dissolution of uranium dioxide to more soluble U(VI) species could occur. This could lead to the mobilization of uranium and other components of spent fuel matrix including fission products and transuranium elements. The reduction of uranium back to oxidation state U(IV) can be considered as a favourable immobilization mechanism in a long-term, leading to precipitation due to the low solubility of U(IV) species. The cast iron insert of the disposal canister and its anaerobic corrosion products are the most important reductants under disposal conditions, but dissolved ferrous iron may also function as reductant. Other iron sources in the buffer or near-field rock, are also considered as possible reductants. The reduction of uranium is a very challenging phenomenon to investigate. The experimental studies need e.g. well-controlled anoxic conditions and measurements of oxidation states. Reduction and other simultaneous phenomena are difficult to distinghuish. The groundwater conditions (pH, Eh and ions) influence on the prevailing complexes of U and Fe and on forming corrosion products of iron and, thus they determine also the redox chemistry. The partial reduction of

Spent nuclear fuel retrieval system fuel handling development testing. Final report

International Nuclear Information System (INIS)

Jackson, D.R.; Meeuwsen, P.V.

1997-09-01

Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems' Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment

TMI-2 fuel-recovery plant. Feasibility study

International Nuclear Information System (INIS)

Evans, D.L.

1982-12-01

This project is a feasibility study for constructing a TMI-2 core Fuel Recovery Plant at the Idaho National Engineering Laboratory (INEL). The primary objectives of the Fuel Recovery Plant (FRP) are to recover and account for the fuel and to process, isolate, and package the waste material from the TMI-2 core. This feasibility study is predicated on a baseline plant and covers its design, fabrication, installation, testing and operation. Alternative methods for the disposal of the TMI-2 core have also been considered, but not examined in detail for their feasibility. The FRP will receive TMI-2 fuel in canisters. The fuel will vary from core debris to intact fuel assemblies and include some core structural materials. The canister contents will be shredded and subsequently fed to a dissolver. Uranium, plutonium, fission products, and some core structural material will be dissolved. The uranium will be separated by solvent extraction and solidified by calcination. The plutonium will also be separated by solvent extraction and routed to the Plutonium Extraction Facility. The wastes will be packaged for further treatment, temporary storage or permanent disposal

Thermal analyses of spent nuclear fuel repository

International Nuclear Information System (INIS)

Ikonen, K.

2003-06-01

This report contains the temperature dimensioning of the KBS-3V type 1- or 2-panel repository based on the rock properties measured from the Olkiluoto investigations. The report describes first the development of a calculation methodology for the thermal analysis of a repository for nuclear fuel. The disposed canisters produce residual heat due to decay (or disintegration) of radioactive products. The decay heat is conducted to surrounding rock mass. The methods were applied to determine the effect of different parameters on the highest canister temperature and to support the planning, dimensioning and operation of the repository. The thermal diffusivity of the rock is low and the heat released from the canisters is spread into the surrounding rock volume quite slowly causing thermal gradient in the rock close to canisters and the canister temperature is increased remarkably. The maximum temperature on the canister surface is limited to the design temperature of +100 deg C. However, due to uncertainties in thermal analysis parameters (like scattering in rock conductivity) the allowable calculated maximum canister temperature is set to 90 deg C causing a safety margin of 10 deg C. The allowable temperature is controlled by the spacing between adjacent canisters, adjacent tunnels and the distance between separate panels of the repository and the pre-cooling time affecting power of the canisters. Because of the fact that the disposal operation takes several decades, the moment of disposal of an individual canister in addition to the location has an influence on the maximum temperature in the canister. Also, a second disposal panel in the repository has a thermal interaction with the other panel. This interaction is expressed after a few decades at the strongest. It became apparent that the temperature of canister surfaces can be determined by analytic line heat source model much more efficiently than by numerical analysis, if the analytic model is first verified and

Development of a facility for fabricating nuclear waste canisters from radioactively contaminated steel

International Nuclear Information System (INIS)

Logan, J.A.; Larsen, M.M.

1986-01-01

This paper describes design of a facility and processes capable of using radioactively contaminated waste steel as the principal raw material for fabricating stainless steel canisters to be used for disposal of nuclear high-level waste. By such action, expenditure (i.e., permanent loss to society) of thousands of tons of uncontaminated chromium and nickel to fabricate such canisters can be avoided. Moreover, the cost and risks involved in disposing of large accumulations of radioactively contaminated steel as low-level radioactive waste (LLRW), that would otherwise be necessary, can also be avoided. The canister fabrication processes (involving centrifugal casting) described herein have been tested and proven for this application. The performance characteristics of stainless steel canisters so fabricated have been tested and agreed to by the organizations that have been involved in this development work (Battelle Memorial Institute, DuPont, EGandG and the Savannah River Laboratory) as equivalent to the performance characteristics of canisters fabricated of uncontaminated wrought stainless steel. It is estimated that the production cost for fabricating canisters by the methods described will not differ greatly from the production cost using uncontaminated wrought steel, and the other costs avoided by not having to dispose of the contaminated steel as LLRW could cause this method to produce the lowest ultimate overall costs

Cost effects of Cu powder and bentonite on the disposal costs of an HLW repository in

International Nuclear Information System (INIS)

Kim, Sung Ki; Lee, Min Soo; Lee, Jong Youl; Choi, Heui Joo; Choi, Jong Won

2008-01-01

This paper provides the cost effect results of Cu powder and bentonite on the disposal cost for an HLW repository in Korea. In the cost analysis for both of these cost drivers, the price of Cu powder and the bentonite can affect the canister cost and the bentonite cost of the disposal holes as well as backfilling cost of the tunnels, respectively. Finally, we found that the unit cost of Cu and bentonite was the dominant cost drivers for the surface and underground facilities of an HLW repository. Therefore, an optimization of a canister and the layout of a disposal hole and disposal tunnels are essential to decrease the direct disposal cost of spent fuels. The disposal costs can be largely divided into two parts such as a surface facilities' cost and an underground facilities' cost. According to the KRS' cost analysis, the encapsulation material as well as the buffering and backfilling cost were the significant costs. Especially, a canister's cost was approximately estimated to be more than one fourth of the overall disposal costs. So it can be estimated that the unit cost of Cu powder is an important cost diver. Because the outer shell of the canister was made of Cu powder by a cold spray coating method. In addition, the unit cost of bentonite can also affect the buffering and the backfilling costs of the disposal holes and the disposal tunnels. But, these material costs will be highly expensive and unstable due to the modernization of the developing countries. So the studies for a material cost should be continued to identify the actual cost of an HLW repository

Quality Assurance Program Plan for Project W-379: Spent Nuclear Fuels Canister Storage Building Projec

International Nuclear Information System (INIS)

Duncan, D.W.

1995-01-01

This document describes the Quality Assurance Program Plan (QAPP) for the Spent Nuclear Fuels (SNF) Canister Storage Building (CSB) Project. The purpose of this QAPP is to control project activities ensuring achievement of the project mission in a safe, consistent and reliable manner

Enhanced Earthquake-Resistance on the High Level Radioactive Waste Canister

International Nuclear Information System (INIS)

Choi, Youngchul; Yoon, Chanhoon; Lee, Jeaowan; Kim, Jinsup; Choi, Heuijoo

2014-01-01

In this paper, the earthquake-resistance type buffer was developed with the method protecting safely about the earthquake. The main parameter having an effect on the earthquake-resistant performance was analyzed and the earthquake-proof type buffer material was designed. The shear analysis model was developed and the performance of the earthquake-resistance buffer material was evaluated. The dynamic behavior of the radioactive waste disposal canister was analyzed in case the earthquake was generated. In the case, the disposal canister gets the serious damage. In this paper, the earthquake-resistance buffer material was developed in order to prevent this damage. By putting the buffer in which the density is small between the canister and buffer, the earthquake-resistant performance was improved about 80%

Thermo-mechanical FE-analysis of butt-welding of a Cu-Fe canister for spent nuclear fuel

International Nuclear Information System (INIS)

Josefson, B.L.; Karlsson, L.; Lindgren, L.E.; Jonsson, M.

1992-10-01

In the Swedish nuclear waste program it has been proposed that spent nuclear fuel shall be placed in composite copper-steel canisters. These canisters will be placed in holes in tunnels located some 500 m underground in a rock storage. The canisters consists of two cylinders of 4850 mm length, one inner cylinder made of steel and one outer cylinder made of copper. The outer diameter of the canister is 880 mm and the wall thickness for each cylinder is 50 mm. At the storage, the steel cylinder, which contains the spent nuclear fuel, is placed inside the copper cylinder. Thereafter, a copper end is butt welded to the copper cylinder using electron beam welding. To obtain penetration through the thickness with this weld method a backing ring is placed at the inside of the copper cylinder. In the present paper, the temperature, strain and stress fields present during welding and after cooling after welding are calculated numerically using the FE-code NIKE-2D. The aim is to use the results of the present calculations to estimate the risk for creep fracture during the subsequent design life. A large strain formulation is employed for the calculation of transient and residual stresses in the canister based on the calculated history of the temperature field present in the canister during the welding process. The contact algorithm available in NIKE-2D is used to detect possible contact between the steel and copper cylinders during the welding. To simplify the numerical calculations and reduce the computational time, rotational symmetry is assumed. For large gap distances between the steel and copper cylinders the residual stress field is calculated to have a shape similar to that observed in butt welded pipes with maximum axial stress values at the yield stress level. For small gap distances the backing ring will come in contact with the steel cylinder which leads to large residual stresses in the backing ring. The maximum accumulated plastic strain in the weld metal and

Encapsulation plant preliminary design, phase 2. Repository connected facility

International Nuclear Information System (INIS)

Kukkola, T.

2006-12-01

The disposal facility of the spent nuclear fuel will be located in Olkiluoto. The encapsulation plant is a part of the disposal facility. In this report, an independent encapsulation plant is located above the underground repository. In the encapsulation plant, the spent fuel is received and treated for disposal. In the fuel handling cell, the spent fuel assemblies are unloaded from the spent fuel transport casks and loaded into the disposal canisters. The gas atmosphere of the disposal canister is changed, the bolted inner canister lid is closed, and the electron beam welding method is used to close the lid of the outer copper canister. The disposal canisters are cleaned and transferred into the buffer store after the machining and inspection of the copper lid welds. From the buffer store, the disposal canisters are transferred into the repository spaces by help of the canister lift. All needed stages of operation are to be performed safely without any activity releases or remarkable personnel doses. The bentonite block interim storage is associated with the encapsulation plant. The bentonite blocks are made from bentonite powder. The bentonite blocks are used as buffer material around the disposal canister in the deposition hole. The average production rate of the encapsulation plant is 40 canisters per year. The nominal maximum production capacity is 100 canisters per year in one shift operation. (orig.)

Neutron interrogator assay system for the Idaho Chemical Processing Plant waste canisters and spent fuel: preliminary description and operating procedures manual

International Nuclear Information System (INIS)

Menlove, H.O.; Eccleston, G.; Close, D.A.; Speir, L.G.

1978-05-01

A neutron interrogation assay system is being designed for the measurement of waste canisters and spent fuel packages at the new Idaho Chemical Processing Plant to be operated by Allied Chemical Corp. The assay samples consist of both waste canisters from the fluorinel dissolution process and spent fuel assemblies. The assay system is a 252 Cf ''Shuffler'' that employs a cyclic sequence of fast-neutron interrogation with a 252 Cf source followed by delayed-neutron counting to determine the 235 U content

Development of Copper Canister through Cold Sprayed Coating Method

Energy Technology Data Exchange (ETDEWEB)

Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo; Lee, Jong Youl; Jeong, Jong Tae; Kim, Sung Ki; Cho, Dong Keun

2007-12-15

General thickness of a copper canister is 5 cm for a underground disposal application. The lower limit of a thickness is determined by a forging technology. But many experts in this area agrees that the thickness 1 cm is enough at the underground disposal for the life time of 1,000,000 years. Thus new technology is suggested for the making 1 cm thickness copper canister, that is a cold spray coating method(CSC). In this report, the CSC is examined and the technical possibility for making copper canister is measured. The overview of CSC and its characteristics are discussed. Various copper particles for the CSC are analyzed and the formed coating layers are examined to find their porosity and uniformity. A Tafa copper particle and Chang-sung copper particle are selected for making 1 cm thick test specimen. Using the CSC specimens, tensile test and XRD analysis are performed. As a corrosion evaluation, a electrochemical test such as a polarization test is done, together with humid corrosion test and chloric acid immersion test. Through the corrosion tests, it is tried to confirm that the CSC is valuable method for making a copper canister. Consequently, it is confirmed that the CSC method is very usful for making 1 cm thick copper canister. the porosity of CSC layer is very low at 0.3 in case of Tafa copper layer. In corrosion tests, the CSC layers are very stable in active environments. It is hard to say that the difference of processing method but the purity of copper is important for the corrosion rate evaluation. The CSC method is very effective method for making 1 cm thick copper canister, It is hoped that the CSC method is applied in a HLW underground disposal system in the future.

Development of Copper Canister through Cold Sprayed Coating Method

International Nuclear Information System (INIS)

Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo; Lee, Jong Youl; Jeong, Jong Tae; Kim, Sung Ki; Cho, Dong Keun

2007-12-01

General thickness of a copper canister is 5 cm for a underground disposal application. The lower limit of a thickness is determined by a forging technology. But many experts in this area agrees that the thickness 1 cm is enough at the underground disposal for the life time of 1,000,000 years. Thus new technology is suggested for the making 1 cm thickness copper canister, that is a cold spray coating method(CSC). In this report, the CSC is examined and the technical possibility for making copper canister is measured. The overview of CSC and its characteristics are discussed. Various copper particles for the CSC are analyzed and the formed coating layers are examined to find their porosity and uniformity. A Tafa copper particle and Chang-sung copper particle are selected for making 1 cm thick test specimen. Using the CSC specimens, tensile test and XRD analysis are performed. As a corrosion evaluation, a electrochemical test such as a polarization test is done, together with humid corrosion test and chloric acid immersion test. Through the corrosion tests, it is tried to confirm that the CSC is valuable method for making a copper canister. Consequently, it is confirmed that the CSC method is very usful for making 1 cm thick copper canister. the porosity of CSC layer is very low at 0.3 in case of Tafa copper layer. In corrosion tests, the CSC layers are very stable in active environments. It is hard to say that the difference of processing method but the purity of copper is important for the corrosion rate evaluation. The CSC method is very effective method for making 1 cm thick copper canister, It is hoped that the CSC method is applied in a HLW underground disposal system in the future

Hanford spent nuclear fuel project update

Energy Technology Data Exchange (ETDEWEB)

Williams, N.H.

1997-08-19

Twenty one hundred metric tons of spent nuclear fuel (SNF) are currently stored in the Hanford Site K Basins near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported to the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building.

DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

International Nuclear Information System (INIS)

BAZINET, G.D.

2003-01-01

The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed

Nuclear fuel waste disposal in Canada

International Nuclear Information System (INIS)

Dormuth, K.W.; Gillespie, P.A.

1990-05-01

Atomic Energy of Canada Limited (AECL) has developed a concept for disposing of Canada's nuclear fuel waste and is submitting it for review under Federal Environmental Assessment and Review Process. During this review, AECL intends to show that careful, controlled burial 500 to 1000 metres deep in plutonic rock of the Canadian Precambrian Shield is a safe and feasible way to dispose of Canada's nuclear fuel waste. The concept has been assessed without identifying or evaluating any particular site for disposal. AECL is now preparing a comprehensive report based on more than 10 years of research and development

Nuclear fuel waste disposal in Canada

International Nuclear Information System (INIS)

Dormuth, K.W.; Gillespie, P.A.

1990-05-01

Atomic Energy of Canada Limited (AECL) has developed a concept for disposing of Canada's nuclear fuel waste and is submitting it for review under the Federal Environmental Assessment and Review Process. During this review, AECL intends to show that careful, controlled burial 500 to 1000 metres deep in plutonic rock of the Canadian Precambrian Shield is a safe and feasible way to dispose of Canada's nuclear fuel waste. The concept has been assessed without identifying or evaluating any particular site for disposal. AECL is now preparing a comprehensive report based on more than 10 years of research and development

Transporting existing VSC-24 canisters using a risk-based licensing approach

Energy Technology Data Exchange (ETDEWEB)

Srinivasan, R.; Sisley, S.E.; Hopf, J.E. [BNFL Fuel Solutions, Campbell, CA (United States)

2004-07-01

The eventual disposition of the spent fuel assemblies loaded in canisters and casks currently designed and licensed only for on-site storage is an industry-wide issue. The canister-specific BUC evaluation approach developed by BFS can be used to license many of these storage canisters and casks for transportation. This will allow these storage canisters and casks to be transported intact to a long-term storage facility or repository, thereby minimizing fuel handling operations, impact on plant operations, and occupational exposure, as well as total infrastructure costs. Application of the proposed canister-specific BUC analysis approach to a preliminary evaluation of the 58 loaded MSBs demonstrates the benefits of this approach. The results of this preliminary evaluation show that a more rigorous analysis based on the known characteristics of the loaded spent fuel, rather than the design-basis fuel parameters, produces significantly lower maximum keff values and can be used to qualify many of the existing loaded storage canisters for transportation. Transportation certification for storage canisters having more reactive spent fuel payloads may require reliance on BUC approaches that are more aggressive than current NRC guidelines allow. Credit may be required for fission- product isotopes that do not have sufficient chemical assay data for benchmarking. In addition, reduced criticality safety margins may be required. For these more-aggressive BUC approaches, a risk assessment should be provided to support the NRC-approval basis. The risk assessment should evaluate the possibility and consequences of an accidental criticality event based upon inaccuracies in the characterization of the spent-fuel payloads.

Transporting existing VSC-24 canisters using a risk-based licensing approach

International Nuclear Information System (INIS)

Srinivasan, R.; Sisley, S.E.; Hopf, J.E.

2004-01-01

The eventual disposition of the spent fuel assemblies loaded in canisters and casks currently designed and licensed only for on-site storage is an industry-wide issue. The canister-specific BUC evaluation approach developed by BFS can be used to license many of these storage canisters and casks for transportation. This will allow these storage canisters and casks to be transported intact to a long-term storage facility or repository, thereby minimizing fuel handling operations, impact on plant operations, and occupational exposure, as well as total infrastructure costs. Application of the proposed canister-specific BUC analysis approach to a preliminary evaluation of the 58 loaded MSBs demonstrates the benefits of this approach. The results of this preliminary evaluation show that a more rigorous analysis based on the known characteristics of the loaded spent fuel, rather than the design-basis fuel parameters, produces significantly lower maximum keff values and can be used to qualify many of the existing loaded storage canisters for transportation. Transportation certification for storage canisters having more reactive spent fuel payloads may require reliance on BUC approaches that are more aggressive than current NRC guidelines allow. Credit may be required for fission- product isotopes that do not have sufficient chemical assay data for benchmarking. In addition, reduced criticality safety margins may be required. For these more-aggressive BUC approaches, a risk assessment should be provided to support the NRC-approval basis. The risk assessment should evaluate the possibility and consequences of an accidental criticality event based upon inaccuracies in the characterization of the spent-fuel payloads

Nuclear fuel waste disposal

International Nuclear Information System (INIS)

Allan, C.J.

1993-01-01

The Canadian concept for nuclear fuel waste disposal is based on disposing of the waste in a vault excavated 500-1000 m deep in intrusive igneous rock of the Canadian Shield. The author believes that, if the concept is accepted following review by a federal environmental assessment panel (probably in 1995), then it is important that implementation should begin without delay. His reasons are listed under the following headings: Environmental leadership and reducing the burden on future generations; Fostering public confidence in nuclear energy; Forestalling inaction by default; Preserving the knowledge base. Although disposal of reprocessing waste is a possible future alternative option, it will still almost certainly include a requirement for geologic disposal

Thermal Predictions of the Cooling of Waste Glass Canisters

Energy Technology Data Exchange (ETDEWEB)

Donna Post Guillen

2014-11-01

Radioactive liquid waste from five decades of weapons production is slated for vitrification at the Hanford site. The waste will be mixed with glass forming additives and heated to a high temperature, then poured into canisters within a pour cave where the glass will cool and solidify into a stable waste form for disposal. Computer simulations were performed to predict the heat rejected from the canisters and the temperatures within the glass during cooling. Four different waste glass compositions with different thermophysical properties were evaluated. Canister centerline temperatures and the total amount of heat transfer from the canisters to the surrounding air are reported.

Synthesis on the spent fuel long term evolution

Energy Technology Data Exchange (ETDEWEB)

Ferry, C.; Poinssot, Ch.; Lovera, P.; Poulesquen, A. [CEA Saclay, Dept. de Physico-Chimie (DEN/DPC), 91 - Gif sur Yvette (France); Broudic, V. [CEA Cadarache, Direction des Reacteurs Nucleaires (DRN), 13 - Saint Paul lez Durance (France); Cappelaere, Ch. [CEA Saclay, Dept. des Materiaux pour le Nucleaire(DMN), 91 - Gif-sur-Yvette (France); Desgranges, L. [CEA Cadarache, Direction des Reacteurs Nucleaires (DRN), 13 - Saint-Paul-lez-Durance (France); Garcia, Ph. [CEA Cadarache, Dept. d' Etudes des Combustibles (DEC), 13 - Saint Paul lez Durance (France); Jegou, Ch.; Roudil, D. [CEA Valrho, Dir. de l' Energie Nucleaire (DEN), 30 - Marcoule (France); Lovera, P.; Poulesquen, A. [CEA Saclay, Dept. de Physico-Chimie (DPC), 91 - Gif sur Yvette (France); Marimbeau, P. [CEA Cadarache, Dir. de l' Energie Nucleaire (DEN), 13 - Saint-Paul-lez-Durance (France); Gras, J.M.; Bouffioux, P. [Electricite de France (EDF), 75 - Paris (France)

2005-07-01

The French research on spent fuel long term evolution has been performed by CEA (Commissariat a l'Energie Atomique) since 1999 in the PRECCI project with the support of EDF (Electricite de France). These studies focused on the spent fuel behaviour under various conditions encountered in dry storage or in deep geological disposal. Three main types of conditions were discerned: - The evolution in a closed system which corresponds to the normal scenario in storage and to the first confinement phase in disposal; - The evolution in air which corresponds to an incidental loss of confinement during storage or to a rupture of the canister before the site re-saturation in geological disposal; - The evolution in water which corresponds to the normal scenario after the breaching of the canister in repository conditions. This document produced in the frame of the PRECCI project is an overview of the state of knowledge in 2004 concerning the long-term behavior of spent fuel under these various conditions. The state of the art was derived from the results obtained under the PRECCI project as well as from a review of the literature and of data acquired under the European project on Spent Fuel Stability under Repository Conditions. The main results issued from the French research are underlined. (authors)

Synthesis on the spent fuel long term evolution

International Nuclear Information System (INIS)

Ferry, C.; Poinssot, Ch.; Lovera, P.; Poulesquen, A.; Broudic, V.; Cappelaere, Ch.; Desgranges, L.; Garcia, Ph.; Jegou, Ch.; Roudil, D.; Lovera, P.; Poulesquen, A.; Marimbeau, P.; Gras, J.M.; Bouffioux, P.

2005-01-01

The French research on spent fuel long term evolution has been performed by CEA (Commissariat a l'Energie Atomique) since 1999 in the PRECCI project with the support of EDF (Electricite de France). These studies focused on the spent fuel behaviour under various conditions encountered in dry storage or in deep geological disposal. Three main types of conditions were discerned: - The evolution in a closed system which corresponds to the normal scenario in storage and to the first confinement phase in disposal; - The evolution in air which corresponds to an incidental loss of confinement during storage or to a rupture of the canister before the site re-saturation in geological disposal; - The evolution in water which corresponds to the normal scenario after the breaching of the canister in repository conditions. This document produced in the frame of the PRECCI project is an overview of the state of knowledge in 2004 concerning the long-term behavior of spent fuel under these various conditions. The state of the art was derived from the results obtained under the PRECCI project as well as from a review of the literature and of data acquired under the European project on Spent Fuel Stability under Repository Conditions. The main results issued from the French research are underlined. (authors)

Life Prediction of Spent Fuel Storage Canister Material

Energy Technology Data Exchange (ETDEWEB)

Ballinger, Ronald

2018-04-16

The original purpose of this project was to develop a probabilistic model for SCC-induced failure of spent fuel storage canisters, exposed to a salt-air environment in the temperature range 30-70°C for periods up to and exceeding 100 years. The nature of this degradation process, which involves multiple degradation mechanisms, combined with variable and uncertain environmental conditions dictates a probabilistic approach to life prediction. A final report for the original portion of the project was submitted earlier. However, residual stress measurements for as-welded and repair welds could not be performed within the original time of the project. As a result of this, a no-cost extension was granted in order to complete these tests. In this report, we report on the results of residual stress measurements.

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

Safeguards considerations related to the use of multi-purpose canisters in the Civilian Radioactive Waste Management system

International Nuclear Information System (INIS)

Floyd, W.C.

1995-01-01

The US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is responsible for disposing of the nation's high-level radioactive waste. Currently, DOE is considering the use of Multi-Purpose Canisters (MPCs) to containerize commercial spent nuclear fuel (SNF) to be handled by the system. To achieve its safeguards and security objectives, OCRWM plans to institute a US Regulatory Commission (NRC)-approved safeguards program. Since the Mined Geologic Disposal System (MGDS) facility and a possible Monitored Retrievable Storage (MRS) facility may be subject to selection for International Atomic Energy Agency (IAEA) inspections, the safeguards program for MPCs may not preclude compliance with the requirements of the IAEA's Annex D, Special Criteria for Difficult-to-Access Fuel Items. MPC safeguards are based on three principles: Verification, Material Control and Accounting, and Physical Protection

Modelling of thermally driven groundwater flow in a facility for disposal of spent nuclear fuel in deep boreholes

Energy Technology Data Exchange (ETDEWEB)

Marsic, Nico; Grundfelt, Bertil [Kemakta Konsult AB, Stockholm (Sweden)

2013-09-15

In this report calculations are presented of buoyancy driven groundwater flow caused by the emission of residual heat from spent nuclear fuel deposited in deep boreholes from the ground surface in combination with the natural geothermal gradient. This work has been conducted within SKB's programme for evaluation of alternative methods for final disposal of spent nuclear fuel. The basic safety feature of disposal of spent nuclear fuel in deep boreholes is that the groundwater at great depth has a higher salinity, and hence a higher density, than more superficial groundwater. The result of this is that the deep groundwater becomes virtually stagnant. The study comprises analyses of the effects of different inter-borehole distances as well as the effect of different permeabilities in the backfill and sealing materials in the borehole and of different shapes of the interface between fresh and saline groundwater. The study is an update of a previous study published in 2006. In the present study, the facility design proposed by Sandia National Laboratories has been studied. In this design, steel canisters containing two BWR elements or one PWR element are stacked on top of each other between 3 and 5 kilometres depth. In order to host all spent fuel from the current Swedish nuclear programme, about 80 such holes are needed. The model used in this study comprises nine boreholes spaced 100 metres alternatively 50 metres apart in a 3{Chi}3 matrix. In one set of calculations the salinity in the groundwater was assumed to increase from zero above 700 metres depth to 10% by weight at 1500 metres depth and below. In another set, a sharper salinity gradient was applied in which the salinity increased from 0 to 10% between 1400 and 1500 metres depth. A geothermal gradient of 16 deg C/km was applied. The heat output from the spent fuel was assumed to decrease by time in manner consistent with the radioactive decay in the fuel. When the inter-borehole distance decreased from

Treatment and disposal of radioactive wastes from nuclear power plants. Program for encapsulation, deep geologic deposition and research, development and demonstration

International Nuclear Information System (INIS)

1995-09-01

Programs for RD and D concerning disposal of radioactive waste are presented. Main topics include: Design, testing and manufacture of canisters for the spent fuels; Design of equipment for deposition of waste canisters; Material and process for backfilling rock caverns; Evaluation of accuracy and validation of methods for safety analyses; Development of methods for defining scenarios for the safety analyses. 471 refs, 67 figs, 21 tabs

Thermal dimensioning of spent fuel repository

International Nuclear Information System (INIS)

Ikonen, K.

2009-09-01

This report contains the temperature dimensioning of the KBS-3V type nuclear fuel repository in Olkiluoto for the BWR, VVER and EPR fuel canisters, which are disposed at vertical position in the horizontal tunnels in a rectangular geometry according to the preliminary Posiva plan. This report concerns only the temperature dimensioning of the repository and does not take into account the possible restrictions caused by the stresses induced in the rock. The maximum temperature on the canister-bentonite interface is limited to the design temperature of +100 deg C. However, due to uncertainties in thermal analysis parameters (like scattering in rock conductivity or in predicted decay power) the allowable calculated maximum canister temperature is set to 90 deg C causing a safety margin of 10 deg C. The allowable temperature is controlled by adjusting the space between adjacent canisters, adjacent tunnels and the pre-cooling time affecting on power of the canisters. The temperature of canister surfaces can be determined by superposing analytic line heat source models much more efficiently than by numerical analysis, if the analytic model is first calibrated by numerical analysis (by control volume method). This was done by comparing the surface temperatures of a single canister calculated numerically and analytically. For the Olkiluoto repository of one panel having 900 canisters of BWR, VVER and EPR spent fuel was analyzed. The analyses were performed with an initial canister power of 1 700 W, 1 370 W and 1 830 W, respectively. These decay heats are obtained when the pre-cooling times of the fuels are 32.9, 29.6 and 50.3 years (the burn-up values 40, 40 and 50 MWd/kgU, respectively). The analyses gave as a result the canister spacing (6.0-10.8 m), when the tunnel spacing was 25 m, 30 m or 40 m. On the edge areas of the panel with constant canister spacing the temperatures of the canisters are lower than in the middle area of the repository. Thus it is possible to pack

State of the art of the welding method for sealing spent nuclear fuel canister made of copper. Part 1 - FSW

Energy Technology Data Exchange (ETDEWEB)

Purhonen, T.

2014-05-15

The purpose of this report is to gather together comprehensive information concerning FSW as an optional welding method for welding the nuclear waste copper canister at the disposal facility. This report discusses the current situation, knowledge of the process and information concerning results of the development and research work related to welding thick copper and the special needs of the disposal environment. Most of the research work and development work has been done by Posiva's Swedish partner SKB, Swedish Nuclear Fuel and Waste Management Co. SKB chose FSW as their reference welding method in 2005. FSW (friction stir welding) is a solid-state welding method, invented in 1991, in which frictional heat is generated between the tool and the weld metal, causing the metal to soften, normally without reaching the melting point, and allowing the tool to traverse the joint line. Friction stir welding can be used for joining many types of materials and material combinations, if the tool materials and designs can be found which operate at the forging temperature of the workpiece. The general requirements for the copper canister weld and base material are presented in Posiva's VAHA-system, which sets the most critical values or demands concerning the short- and long-term properties or other needs. The sections in this report are set out in a similar way as in the VAHA-system. Concerning the results from the research and development work, it can be said that FS weld material fulfils the values set by VAHA. The quality of the welds fulfils the set demands for intact weld material and the welding process is robust using an automatic control system. There still remains work concerning the acceptance procedure for the welding process and other open issues which are described in this report. (orig.)

State of the art of the welding method for sealing spent nuclear fuel canister made of copper. Part 1 - FSW

International Nuclear Information System (INIS)

Purhonen, T.

2014-05-01

The purpose of this report is to gather together comprehensive information concerning FSW as an optional welding method for welding the nuclear waste copper canister at the disposal facility. This report discusses the current situation, knowledge of the process and information concerning results of the development and research work related to welding thick copper and the special needs of the disposal environment. Most of the research work and development work has been done by Posiva's Swedish partner SKB, Swedish Nuclear Fuel and Waste Management Co. SKB chose FSW as their reference welding method in 2005. FSW (friction stir welding) is a solid-state welding method, invented in 1991, in which frictional heat is generated between the tool and the weld metal, causing the metal to soften, normally without reaching the melting point, and allowing the tool to traverse the joint line. Friction stir welding can be used for joining many types of materials and material combinations, if the tool materials and designs can be found which operate at the forging temperature of the workpiece. The general requirements for the copper canister weld and base material are presented in Posiva's VAHA-system, which sets the most critical values or demands concerning the short- and long-term properties or other needs. The sections in this report are set out in a similar way as in the VAHA-system. Concerning the results from the research and development work, it can be said that FS weld material fulfils the values set by VAHA. The quality of the welds fulfils the set demands for intact weld material and the welding process is robust using an automatic control system. There still remains work concerning the acceptance procedure for the welding process and other open issues which are described in this report. (orig.)

DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

Energy Technology Data Exchange (ETDEWEB)

BAZINET, G.D.

2003-02-12

The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed.

Disposal of Surplus Weapons Grade Plutonium

International Nuclear Information System (INIS)

Alsaed, H.; Gottlieb, P.

2000-01-01

The Office of Fissile Materials Disposition is responsible for disposing of inventories of surplus US weapons-usable plutonium and highly enriched uranium as well as providing, technical support for, and ultimate implementation of, efforts to obtain reciprocal disposition of surplus Russian plutonium. On January 4, 2000, the Department of Energy issued a Record of Decision to dispose of up to 50 metric tons of surplus weapons-grade plutonium using two methods. Up to 17 metric tons of surplus plutonium will be immobilized in a ceramic form, placed in cans and embedded in large canisters containing high-level vitrified waste for ultimate disposal in a geologic repository. Approximately 33 metric tons of surplus plutonium will be used to fabricate MOX fuel (mixed oxide fuel, having less than 5% plutonium-239 as the primary fissile material in a uranium-235 carrier matrix). The MOX fuel will be used to produce electricity in existing domestic commercial nuclear reactors. This paper reports the major waste-package-related, long-term disposal impacts of the two waste forms that would be used to accomplish this mission. Particular emphasis is placed on the possibility of criticality. These results are taken from a summary report published earlier this year

Spent fuel disposal problem in Bulgaria

Energy Technology Data Exchange (ETDEWEB)

Milanov, M; Stefanova, I [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

1994-12-31

The internationally agreed basic safety principles and criteria for spent fuel (SF) and high level waste (HLW) disposal are outlined. In the framework of these principles the specific problems of Bulgaria described in the `National Concept for Radioactive Waste Management and Disposal in Republic of Bulgaria` are discussed. The possible alternatives for spent fuel management are: (1) sending the spent fuel for disposal in other country; (2) once-through cycle and (3) closed fuel cycle. A mixed solution of the problem is implemented in Bulgaria. According to the agreement between Bulgaria and former Soviet Union a part of the spent fuel has been returned to Russia. The once-through and closed-fuel cycle are also considered. The projected cumulated amount of spent fuel is estimated for two cases: (1) the six units of Kozloduy NPP are in operation till the end of their lifetime (3300 tHM) and (2) low estimate (2700 tHM) - only units 5 and 6 are operated till the end of their lifetime. The reprocessing of the total amount of 3300 tHM will lead to the production of about 370 m{sup 3} vitrified high level wastes. Together with the HLW about 1850 m{sup 3} cladding hulls and 7800 m{sup 3} intermediate-level wastes will be generated, which should be disposed off in deep geological repository. The total production of radioactive waste in once-through cycle is 181 000 m{sup 3}, and in closed cycle - 190 000 m{sup 3}. Geological investigations are performed resulting in categorization of the territory of the country based on geological, geotechnical and hydrogeological conditions. This will facilitate the choice of the most suitable location for deep geological repository. 7 figs., 11 refs.

Commercial radioactive waste management system feasibility with the universal canister concept. Volume 1

International Nuclear Information System (INIS)

Morissette, R.P.; Schneringer, P.E.; Lane, R.K.; Moore, R.L.; Young, K.A.

1986-01-01

A Program Research and Development Announcement (PRDA) was initiated by DOE to solicit from industry new and novel ideas for improvements in the nuclear waste management system. GA Technologies Inc. was contracted to study a system utilizing a universal canister which could be loaded at the reactor and used throughout the waste management system. The proposed canister was developed with the objective of meeting the mission requirements with maximum flexibility and at minimum cost. Canister criteria were selected from a thorough analysis of the spent fuel inventory, and canister concepts were evaluated along with the shipping and storage casks to determine the maximum payload. Engineering analyses were performed on various cask/canister combinations. One important criterion was the interchangeability of the canisters between truck and rail cask systems. A canister was selected which could hold three PWR intact fuel elements or up to eight consolidated PWR fuel elements. One canister could be shipped in an overweight truck cask or six in a rail cask. Economic analysis showed a cost savings of the reference system under consideration at that time

Draft Geologic Disposal Requirements Basis for STAD Specification

Energy Technology Data Exchange (ETDEWEB)

Ilgen, Anastasia G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-03-25

This document provides the basis for requirements in the current version of Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems, (FCRD-NFST-2014-0000579) that are driven by storage and geologic disposal considerations. Performance requirements for the Standardized Transportation, Aging, and Disposal (STAD) canister are given in Section 3.1 of that report. Here, the requirements are reviewed and the rationale for each provided. Note that, while FCRD-NFST-2014-0000579 provides performance specifications for other components of the STAD storage system (e.g. storage overpack, transfer and transportation casks, and others), these have no impact on the canister performance during disposal, and are not discussed here.

Hanford K basins spent nuclear fuel project update

International Nuclear Information System (INIS)

Williams, N.H.; Hudson, F.G.

1997-07-01

Twenty one hundred metric tons of spent nuclear fuel (SNF) are currently stored in the Hanford Site K Basins near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported to the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building

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

Radioactive waste disposal package

Science.gov (United States)

Lampe, Robert F.

1986-11-04

A radioactive waste disposal package comprising a canister for containing vitrified radioactive waste material and a sealed outer shell encapsulating the canister. A solid block of filler material is supported in said shell and convertible into a liquid state for flow into the space between the canister and outer shell and subsequently hardened to form a solid, impervious layer occupying such space.

Friction Stir Welding of Copper Canisters for Nuclear Waste

Energy Technology Data Exchange (ETDEWEB)

Kaellgren, Therese

2005-07-01

The Swedish model for final disposal of nuclear fuel waste is based on copper canisters as a corrosion barrier with an inner pressure holding insert of cast iron. One of the methods to seal the copper canister is to use the Friction Stir Welding (FSW), a method invented by The Welding Institute (TWI). This work has been focused on characterisation of the FSW joints, and modelling of the process, both analytically and numerically. The first simulations were based on Rosenthal's analytical medium plate model. The model is simple to use, but has limitations. Finite element models were developed, initially with a two-dimensional geometry. Due to the requirements of describing both the heat flow and the tool movement, three-dimensional models were developed. These models take into account heat transfer, material flow, and continuum mechanics. The geometries of the models are based on the simulation experiments carried out at TWI and at Swedish Nuclear Fuel Waste and Management Co (SKB). Temperature distribution, material flow and their effects on the thermal expansion were predicted for a full-scale canister and lid. The steady state solutions have been compared with temperature measurements, showing good agreement. Microstructure and hardness profiles have been investigated by optical microscope, Scanning Electron Microscope (SEM), Electron Back Scatter Diffraction (EBSD) and Rockwell hardness measurements. EBSD visualisation has been used to determine the grain size distribution and the appearance of twins and misorientation within grains. The orientation maps show a fine uniform equiaxed grain structure. The root of the weld exhibits the smallest grains and many annealing twins. This may be due to deformation after recrystallisation. The appearance of the nugget and the grain size depends on the position of the weld. A large difference can be seen both in hardness and grain size between the start of the weld and when the steady state is reached.

The psychosocial consequences of spent fuel disposal

International Nuclear Information System (INIS)

Paavola, J.; Eraenen, L.

1999-03-01

In this report the potential psychosocial consequences of spent fuel disposal to inhabitants of a community are assessed on the basis of earlier research. In studying the situation, different interpretations and meanings given to nuclear power are considered. First, spent fuel disposal is studied as fear-arousing and consequently stressful situation. Psychosomatic effects of stress and coping strategies used by an individual are presented. Stress as a collective phenomenon and coping mechanisms available for a community are also assessed. Stress reactions caused by natural disasters and technological disasters are compared. Consequences of nuclear power plant accidents are reviewed, e.g. research done on the accident at Three Mile Island power plant. Reasons for the disorganising effect on a community caused by a technological disaster are compared to the altruistic community often seen after natural disasters. The potential reactions that a spent fuel disposal plant can arouse in inhabitants are evaluated. Both short-term and long-term reactions are evaluated as well as reactions under normal functioning, after an incident and as a consequence of an accident. Finally an evaluation of how the decision-making system and citizens' opportunity to influence the decision-making affect the experience of threat is expressed. As a conclusion we see that spent fuel disposal can arouse fear and stress in people. However, the level of the stress is probably low. The stress is at strongest at the time of the starting of the spent fuel disposal plant. With time people get used to the presence of the plant and the threat experienced gets smaller. (orig.)

Canister disposition plan for the DWPF Startup Test Program

International Nuclear Information System (INIS)

Harbour, J.R.; Payne, C.H.

1990-01-01

This report details the disposition of canisters and the canistered waste forms produced during the DWPF Startup Test Program. The six melter campaigns (DWPF Startup Tests FA-13, WP-14, WP-15, WP-16, WP-17, and FA-18) will produce 126 canistered waste forms. In addition, up to 20 additional canistered waste forms may be produced from glass poured during the transition between campaigns. In particular, this canister disposition plan (1) assigns (by alpha-numeric code) a specific canister to each location in the six campaign sequences, (2) describes the method of access for glass sampling on each canistered waste form, (3) describes the nature of the specific tests which will be carried out, (4) details which tests will be carried out on each canistered waste form, (5) provides the sequence of these tests for each canistered waste form, and (6) assigns a storage location for each canistered waste form. The tests are designed to provide evidence, as detailed in the Waste Form Compliance Plan (WCP 1 ), that the DWPF product will comply with the Waste Acceptance Product Specifications (WAPS 2 ). The WAPS must be met before the canistered waste form is accepted by DOE for ultimate disposal at the Federal Repository. The results of these tests will be included in the Waste Form Qualification Report (WQR)

Progress in the understanding of the long-term corrosion behaviour of copper canisters

Science.gov (United States)

King, Fraser; Lilja, Christina; Vähänen, Marjut

2013-07-01

Copper has been proposed as a canister material for the disposal of spent nuclear fuel in a deep geologic repository in a number of countries worldwide. Since it was first proposed for this purpose in 1978, a significant number of studies have been performed to assess the corrosion performance of copper under repository conditions. These studies are critically reviewed and the suitability of copper as a canister material for nuclear waste is re-assessed. Over the past 30-35 years there has been considerable progress in our understanding of the expected corrosion behaviour of copper canisters. Crucial to this progress has been the improvement in the understanding of the nature of the repository environment and how it will evolve over time. With this improved understanding, it has been possible to predict the evolution of the corrosion behaviour from the initial period of warm, aerobic conditions in the repository to the long-term phase of cool, anoxic conditions dominated by the presence of sulphide. An historical review of the treatment of the corrosion behaviour of copper canisters is presented, from the initial corrosion assessment in 1978, through a major review of the corrosion behaviour in 2001, through to the current level of understanding based on the results of on-going studies. Compared with the initial corrosion assessment, there has been considerable progress in the treatment of localised corrosion, stress corrosion cracking, and microbiologically influenced corrosion of the canisters. Progress in the mechanistic modelling of the evolution of the corrosion behaviour of the canister is also reviewed, as is the continuing debate about the thermodynamic stability of copper in pure water. The overall conclusion of this critical review is that copper is a suitable material for the disposal of spent nuclear fuel and offers the prospect of containment of the waste for an extended period of time. The corrosion behaviour is influenced by the presence of the

Utilization of Cs137 to generate a radiation barrier for weapons grade plutonium immobilized in borosilicate glass canisters. Revision 1

International Nuclear Information System (INIS)

Jardine, L.J.; Armantrout, G.A.; Collins, L.F.

1995-01-01

One of the ways recommended by a recent National Academy of Sciences study to dispose of excess weapons-grade plutonium is to encapsulate the plutonium in a glass in combination with high-level radioactive wastes (HLW) to generate an intense radiation dose rate field. The objective is to render the plutonium as difficult to access as the plutonium contained in existing US commercial spent light-water reactor (LWR) fuel until it can be disposed of in a permanent geological repository. A radiation dose rate from a sealed canister of 1,000 rem/h (10 Sv/h) at 1 meter for at least 30 years after fabrication was assumed in this paper to be a radiation dose comparable to spent LWR fuel. This can be achieved by encapsulating the plutonium in a borosilicate glass with an adequate amount of a single fission product in the HLWS, namely radioactive Cs 137 . One hundred thousand curies of Cs 137 will generate a dose rate of 1,000 rem/h (10 Sv/h) at 1 meter for at least 30 years when imbedded into canisters of the size proposed for the Savannah River Site's vitrified high-level wastes. The United States has a current inventory of 54 MCi of CS 137 that has been separated from defense HLWs and is in sealed capsules. This single curie inventory is sufficient to spike 50 metric tons of excess weapons-grade plutonium if plutonium can be loaded at 5.5 wt% in glass, or 540 canisters. Additional CS 137 inventories exist in the United States' HLWs from past reprocessing operations, should additional curies be required. Using only one fission product, CS 137 , rather than the multiple chemical elements and compounds in HLWs to generate a high radiation dose rate from a glass canister greatly simplifies the processing engineering retirement for encapsulating plutonium in a borosilicate glass

Nuclear fuel waste disposal

International Nuclear Information System (INIS)

Merrett, G.J.; Gillespie, P.A.

1983-07-01

This report discusses events and processes that could adversely affect the long-term stability of a nuclear fuel waste disposal vault or the regions of the geosphere and the biosphere to which radionuclides might migrate from such a vault

A study on the thermal design of the advanced multi purpose canister

International Nuclear Information System (INIS)

Jun, Eun Ju; Chang, S. H.

2004-01-01

The storage, transportation and disposal of pressurized water reactor (PWR) spent fuel is an important issue in nuclear industry for safe and extended nuclear power plant operation. However, since wet-storage facilities almost reach to their capacity, dry-storage method, especially using a Multi Purpose Canister (MPC), has been highlighted because it gives better economic and safety point of view. In this study, an advanced design of MPC was proposed through an increase in storage capacity from 26 to 37 PWR fuel assemblies per MPC. Thermal analysis was also carried out by FLUENT 6 code to confirm the temperature criteria of the MPC. From the results, conceptual design of 37 PWR fuel assemblies can reduce the mass per fuel assembly ratio about 19 % compared to the 26 assemblies, and maximum temperature developed at the center of fuel assemblies shows within the range of design criteria. Based on the analysis, it is considered that the advanced MPC design well conforms to the safety requirement and it gives rise to economical gain because of reduction in mass per fuel assembly ratio

Mechanical integrity of canisters

International Nuclear Information System (INIS)

Nilsson, Fred

1992-12-01

This document constitutes the final report from 'SKBs reference group for mechanical integrity of canisters for spent nuclear fuel'. A complete list of all reports initiated by the reference group can be found in the summary report in this document. The main task of the reference group has been to advice SKB regarding the choice (ranking of alternatives) of canister type for different types of storage. The choice should be based on requirements of impermeability for a given time period and identification of possible limiting mechanisms. The main conclusions from the work were: From mechanical point of view, low phosphorous oxygen free copper (Cu-OFP) is a preferred canisters material. It exhibits satisfactory ductility both during tensile and creep testing. The residual stresses in the canisters are of such a magnitude that the estimated time to creep rupture with the data obtained for the Cu-OFP material is essentially infinite. Based on the present knowledge of stress corrosion cracking of copper there appears to be a small risk for such to occur in the projected environment. This risk need some further study. Rock shear movements of the size of 10 cm should pose no direct threat to the integrity of the canisters. Considering mechanical integrity, the composite copper/steel canister is an advantageous alternative. The recommendations for further research included continued studies of the creep properties of copper and of stress corrosion cracking. However, the studies should focus more directly on the design and fabrication aspect of the canister

Interaction between rock, bentonite buffer and canister. FEM calculations of some mechanical effects on the canister in different disposal concepts

International Nuclear Information System (INIS)

Boergesson, L.

1992-07-01

An important task of the buffer of highly compacted bentonite is to offer a mechanical protection to the canister. This role has been investigated by a number of finite element calculations using the complex elasto plastic material models for the bentonite that have been developed on the basis of laboratory tests and adapted to the code ABAQUS. The following main functions and scenarios have been investigated for some different canister types and repository concepts: - The effect of the water and swelling pressure, - The effect of a rock shear perpendicular to the canister axis, - The effect of creep in the copper after a rock shear displacement, - The thermomechanical effects when an initially saturated buffer is used

Fuel corrosion processes under waste disposal conditions

International Nuclear Information System (INIS)

Shoesmith, D.W.

2000-01-01

The release of the majority of radionuclides from spent nuclear fuel under permanent disposal conditions will be controlled by the rate of dissolution of the UO 2 fuel matrix. In this manuscript the mechanism of the coupled anodic (fuel dissolution) and cathodic (oxidant reduction) reactions which constitute the overall fuel corrosion process is reviewed, and the many published observations on fuel corrosion under disposal conditions discussed. The primary emphasis is on summarizing the overall mechanistic behaviour and establishing the primary factors likely to control fuel corrosion. Included are discussions on the influence of various oxidants including radiolytic ones, pH, temperature, groundwater composition, and the formation of corrosion product deposits. The relevance of the data recorded on unirradiated UO 2 to the interpretation of spent fuel behaviour is included. Based on the review, the data used to develop fuel corrosion models under the conditions anticipated in Yucca Mountain (NV, USA) are evaluated

Chemical stability of copper-canisters in deep repository

International Nuclear Information System (INIS)

Ahonen, L.

1995-12-01

The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper-iron canisters and placed deep into the bedrock. The copper wall of the canister provides a long-time shield against corrosion, preventing the high-level nuclear fuel from contact with ground water. In the report, stability of metallic copper and its possible corrosion reactions in the conditions of deep bedrock are evaluated by means of thermo-dynamic calculations. (90 refs., 28 figs., 11 tabs.)

Site-to-canister scale flow and transport in Haestholmen, Kivetty, Olkiluoto and Romuvaara

Energy Technology Data Exchange (ETDEWEB)

Poteri, A.; Laitinen, M. [VTT Energy, Espoo (Finland)

1999-05-01

Radioactive waste is originating from production of electricity in nuclear power plants. Most of the waste has only low or intermediate levels of radioactivity. However, the spent nuclear fuel is highly radioactive and it has to be isolated from the biosphere. The current nuclear waste management plan in Finland is based on direct disposal of the spent nuclear fuel deep underground. The only feasible mechanism for the radionuclides to escape from an underground repository is to be carried by the groundwater flow after the failure of waste containers. The scope of this study is to examine the groundwater flow situation and transport properties in the vicinity of the disposal canister and along the potential release paths from the repository into the biosphere. The results of this study are further applied in the site specific safety analysis of a spent fuel repository. Synthesis is made of the porous medium estimates of the groundwater flow in the regional and site scales and the detailed fracture network analysis of the flow in the canister scale. This synthesis includes estimation of the transport properties from the canister into the biosphere and flow rates around the deposition holes of the waste canisters. The modelling has been carried out for four different sites: Hastholmen, Kivetty, Olkiluoto and Romavaara. According to the simulations groundwater flow rate around the deposition holes is less than about 1 litre/a for about 75 % of the deposition holes. For about 5 % of the deposition holes the flow rates are a few litres per year or higher. The highest flow rates resulted at Hastholmen, in fresh water conditions 10 000 years after present, and at Kivetty. The transport resistances were calculated for the `worst` flow paths that might have impact on the safety of the repository. The total transport resistances from the repository into the biosphere along those flow paths varied between about 40 000 a/m and 5-10{sup 6} a/m. Most of the total transport

Analysis of scenarios for the direct disposal of spent nuclear fuel disposal conditions as expected in Germany

International Nuclear Information System (INIS)

Ashton, P.; Mehling, O.; Mohn, R.; Wingender, H.J.

1990-01-01

This report contains an investigation of aspects of the waste management of spent light water reactor fuel by direct disposal in a deep geological formation on land. The areas covered are: interim dry storage of spent fuel with three options of pre-conditioning; conditioning of spent fuel for final disposal in a salt dome repository; disposal of spent fuel (heat-generating waste) in a salt dome repository; disposal of medium and low-level radioactive wastes in the Konrad mine. Dose commitments, effluent discharges and potential incidents were not found to vary significantly for the various conditioning options/salt dome repository types. Due to uncertainty in the cost estimates, in particular the disposal cost estimates, the variation between the three conditioning options examined is not considered as being significant. The specific total costs for the direct disposal strategy are estimated to lie in the range ECU 600 to 700 per kg hm (basis 1988)

SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB) MULTI CANISTER OVERPACK (MCO) SAMPLING SYSTEM VALIDATION (OCRWM)

International Nuclear Information System (INIS)

BLACK, D.M.; KLEM, M.J.

2003-01-01

Approximately 400 Multi-canister overpacks (MCO) containing spent nuclear fuel are to be interim stored at the Canister Storage Building (CSB). Several MCOs (monitored MCOs) are designated to be gas sampled periodically at the CSB sampling/weld station (Bader 2002a). The monitoring program includes pressure, temperature and gas composition measurements of monitored MCOs during their first two years of interim storage at the CSB. The MCO sample cart (CART-001) is used at the sampling/weld station to measure the monitored MCO gas temperature and pressure, obtain gas samples for laboratory analysis and refill the monitored MCO with high purity helium as needed. The sample cart and support equipment were functionally and operationally tested and validated before sampling of the first monitored MCO (H-036). This report documents the results of validation testing using training MCO (TR-003) at the CSB. Another report (Bader 2002b) documents the sample results from gas sampling of the first monitored MCO (H-036). Validation testing of the MCO gas sampling system showed the equipment and procedure as originally constituted will satisfactorily sample the first monitored MCO. Subsequent system and procedural improvements will provide increased flexibility and reliability for future MCO gas sampling. The physical operation of the sampling equipment during testing provided evidence that theoretical correlation factors for extrapolating MCO gas composition from sample results are unnecessarily conservative. Empirically derived correlation factors showed adequate conservatism and support use of the sample system for ongoing monitored MCO sampling

Design, production and initial state of the canister

International Nuclear Information System (INIS)

Cederqvist, Lars; Johansson, Magnus; Leskinen, Nina; Ronneteg, Ulf

2010-12-01

The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility.The report provides input on the initial state of the canisters to the assessment of the long-term safety, SR-Site. The initial state refers to the properties of the engineered barriers once they have been finally placed in the KBS-3 repository and will not be further handled within the repository facility. In addition, the report provides input to the operational safety report, SR-Operation, on how the canisters shall be handled and disposed. The report presents the design premises and reference design of the canister and verifies the conformity of the reference design to the design premises. The production methods and the ability to produce canisters according to the reference design are described. Finally, the initial state of the canisters and their conformity to the reference design and design premises are presented

Design, production and initial state of the canister

Energy Technology Data Exchange (ETDEWEB)

Cederqvist, Lars; Johansson, Magnus; Leskinen, Nina; Ronneteg, Ulf

2010-12-15

The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility.The report provides input on the initial state of the canisters to the assessment of the long-term safety, SR-Site. The initial state refers to the properties of the engineered barriers once they have been finally placed in the KBS-3 repository and will not be further handled within the repository facility. In addition, the report provides input to the operational safety report, SR-Operation, on how the canisters shall be handled and disposed. The report presents the design premises and reference design of the canister and verifies the conformity of the reference design to the design premises. The production methods and the ability to produce canisters according to the reference design are described. Finally, the initial state of the canisters and their conformity to the reference design and design premises are presented

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

Managing Spent Nuclear Fuel at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Thomas Hill; Denzel L. Fillmore

2005-10-01

The Idaho National Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy derives from the history of the INL as the National Reactor Testing Station, and from its mission to recover HEU from SNF and to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facilities, some 50 years old. SNF at INL has many forms—from intact assemblies down to metallurgical mounts, and some fuel has been wet stored for over 40 years. SNF is stored bare or in metal cans under water, or dry in vaults, caissons or casks. Inspection shows varying corrosion and degradation of the SNF and its storage cans. SNF has been stored in 10 different facilities: 5 pools, one cask storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The pools range in age from 40 years old to the most modern in the US Department of Energy (DOE) complex. The near-term objective is to move SNF from older pools to interim dry storage, allowing shutdown and decommissioning of the older facilities. This move involves drying methods that are dependent on fuel type. The long-term objective is to have INL SNF in safe dry storage and ready to be shipped to the National Repository. The unique features of the INL SNF requires special treatments and packaging to meet the proposed repository acceptance criteria and SNF will be repackaged in standardized canisters for shipment and disposal in the National Repository. Disposal will use the standardized canisters that can be co-disposed with High Level Waste glass logs to limit the total fissile material in a repository waste package. The DOE standardized canister also simplifies the repository handling of the multitude of DOE SNF sizes and shapes.

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

TMI-2 [Three Mile Island Nuclear Power Station] fuel canister and core sample handling equipment used in INEL hot cells

International Nuclear Information System (INIS)

McConnell, J.W. Jr.; Shurtliff, W.T.; Lynch, R.J.; Croft, K.M.; Whitmill, L.J.; Allen, S.M.

1987-01-01

This paper describes the specialized remote handling equipment developed and used at the Idaho National Engineering Laboratory (INEL) to handle samples obtained from the core of the damaged Unit 2 reactor at Three Mile Island Nuclear Power Station (TM-2). Samples of the core were removed, placed in TMI-2 fuel canisters, and transported to the INEL. Those samples will be examined as part of the analysis of the TMI-2 accident. The equipment described herein was designed for removing sample materials from the fuel canisters, assisting with initial examination, and processing samples in preparation for detailed examinations. The more complex equipment used microprocessor remote controls with electric motor drives providing the required force and motion capabilities. The remaining components were unpowered and manipulator assisted

Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

International Nuclear Information System (INIS)

D.R. Jackson; G.R. Kiebel

1999-01-01

Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training

Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

Energy Technology Data Exchange (ETDEWEB)

D.R. Jackson; G.R. Kiebel

1999-08-24

Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

INERT-MATRIX FUEL: ACTINIDE ''BURNING'' AND DIRECT DISPOSAL

International Nuclear Information System (INIS)

Rodney C. Ewing; Lumin Wang

2002-01-01

Excess actinides result from the dismantlement of nuclear weapons (Pu) and the reprocessing of commercial spent nuclear fuel (mainly 241 Am, 244 Cm and 237 Np). In Europe, Canada and Japan studies have determined much improved efficiencies for burnup of actinides using inert-matrix fuels. This innovative approach also considers the properties of the inert-matrix fuel as a nuclear waste form for direct disposal after one-cycle of burn-up. Direct disposal can considerably reduce cost, processing requirements, and radiation exposure to workers

The back end of the fuel cycle and CANDU

International Nuclear Information System (INIS)

Allan, C.J.; Dormuth, K.W.

2001-01-01

CANDU reactor operators have benefited from several advantages of the CANDU system and from AECL's experience, with regard to spent fuel handling, storage and disposal. AECL has over 20 years experience in development and application of medium-term storage and research and development on the disposal of used fuel. As a result of AECL's experience, short-term and medium-term storage and the associated handling of spent CANDU fuel are well proven and economic, with an extremely high degree of public and environmental protection. In fact, both short-term (water-pool) and medium-term (dry canister) storage of CANDU fuel are comparable or lower in cost per unit of energy than for PWRs. Both pool storage and dry spent fuel storage are fully proven, with many years of successful, safe operating experience. AECL's extensive R and D on the permanent disposal of spent-fuel has resulted in a defined concept for Canadian fuel disposal in crystalline rock. This concept was recently confirmed as ''technically acceptable'' by an independent environmental review panel. Thus, the Canadian program represents an international demonstration of the feasibility and safety of geological disposal of nuclear fuel waste. Much of the technology behind the Canadian concept can be adapted to permanent land-based disposal strategies chosen by other countries. In addition, the Canadian development has established a baseline for CANDU fuel permanent disposal costs. Canadian and international work has shown that the cost of permanent CANDU fuel disposal is similar to the cost of LWR fuel disposal per unit of electricity produced. (author)

Integrity of copper/steel canisters under crystalline bedrock repository conditions

International Nuclear Information System (INIS)

Bowyer, W.H.; Sjoblom, R.; Trolle, M.

1996-01-01

In the Swedish nuclear waste disposal programme, the need to store the spent nuclear fuel safely for very long times has prompted a strategy which includes a long life canister. Technical as well as economical considerations related to design, choice of materials and manufacturing technology have lead to the selection of a reference design to be used for the continued development work. The canisters are cylindrical with a diameter close to 1 meter and a height of about 5 meters. In order to meet the need for an appropriate combination of mechanical strength, toughness, durability and corrosion resistance, the canisters comprise an inner vessel made of steel or cast iron to cope with mechanical stresses and an outer vessel made of almost pure copper to provide corrosion resistance. The Swedish nuclear industry has recently extended its development work to full-scale tests. Such experience is needed not least for the evaluation of the long-term integrity of the canister. This work has been closely followed by the Swedish Nuclear Power Inspectorate (SKI) who have also carried out independent investigations and analyses. It should be emphasized that the findings relate to a canister which is under development and cannot, in general, be expected to be relevant for the fully developed canister. Significant results of the analyses include the identification of conceivable modes of canister failures. Such failures may be related to defects, segregation, limitations in inspectability, long term creep properties, adverse mechanical load situations, etc. It is assessed that the distribution functions of these failures might have their largest uncertainties at the tails extending to comparatively short times. Specific issues related to canister manufacture, scaling and non destructive testing which have been found to warrant further investigation are: defects in the copper ingot which may transfer to the rolled copper plate; the amount of work applied during the rolling or

Conceptual designs of radioactive canister transporters

International Nuclear Information System (INIS)

1978-02-01

This report covers conceptual designs of transporters for the vertical, horizontal, and inclined installation of canisters containing spent-fuel elements, high-level waste, cladding waste, and intermediate-level waste (low-level waste is not discussed). Included in the discussion are cask concepts; transporter vehicle designs; concepts for mechanisms for handling and manipulating casks, canisters, and concrete plugs; transporter and repository operating cycles; shielding calculations; operator radiation dosages; radiation-resistant materials; and criteria for future design efforts

Korean Reference HLW Disposal System

Energy Technology Data Exchange (ETDEWEB)

Choi, Heui Joo; Lee, J. Y.; Kim, S. S. (and others)

2008-03-15

This report outlines the results related to the development of Korean Reference Disposal System for High-level radioactive wastes. The research has been supported around for 10 years through a long-term research plan by MOST. The reference disposal method was selected via the first stage of the research during which the technical guidelines for the geological disposal of HLW were determined too. At the second stage of the research, the conceptual design of the reference disposal system was made. For this purpose the characteristics of the reference spent fuels from PWR and CANDU reactors were specified, and the material and specifications of the canisters were determined in term of structural analysis and manufacturing capability in Korea. Also, the mechanical and chemical characteristics of the domestic Ca-bentonite were analyzed in order to supply the basic design parameters of the buffer. Based on these parameters the thermal and mechanical analysis of the near-field was carried out. Thermal-Hydraulic-Mechanical behavior of the disposal system was analyzed. The reference disposal system was proposed through the second year research. At the final third stage of the research, the Korean Reference disposal System including the engineered barrier, surface facilities, and underground facilities was proposed through the performance analysis of the disposal system.

Canister storage building design basis accident analysis documentation

International Nuclear Information System (INIS)

KOPELIC, S.D.

1999-01-01

This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

Seabed disposal of high-level nuclear wastes: an alternative viewpoint

International Nuclear Information System (INIS)

Glasby, G.P.

1985-01-01

Various comments on a published article on subseabed disposal of nuclear wastes are presented. These include the scale of the proposed operation, the technical problems of canister retrievability, the feasibility of the free-fall penetrometer disposal method, canister lifetime, the possible contravention of the 1972 London Dumping Convention and land-based geological repositories as an alternative method of disposal. (author)

Disposal of Canada's nuclear fuel waste

International Nuclear Information System (INIS)

Dormuth, K.W.; Nuttall, K.

1994-01-01

In 1978, the governments of Canada and Ontario established the Nuclear Fuel Waste Management program. As of the time of the conference, the research performed by AECL was jointly funded by AECL and Ontario Hydro through the CANDU owners' group. Ontario Hydro have also done some of the research on disposal containers and vault seals. From 1978 to 1992, AECL's research and development on disposal cost about C$413 million, of which C$305 was from funds provided to AECL by the federal government, and C$77 million was from Ontario Hydro. The concept involves the construction of a waste vault 500 to 1000 metres deep in plutonic rock of the Canadian Precambrian Shield. Used fuel (or possibly solidified reprocessing waste) would be sealed into containers (of copper, titanium or special steel) and emplaced (probably in boreholes) in the vault floor, surrounded by sealing material (buffer). Disposal rooms might be excavated on more than one level. Eventually all excavated openings in the rock would be backfilled and sealed. Research is organized under the following headings: disposal container, waste form, vault seals, geosphere, surface environment, total system, assessment of environmental effects. A federal Environmental Assessment Panel is assessing the concept (holding public hearings for the purpose) and will eventually make recommendations to assist the governments of Canada and Ontario in deciding whether to accept the concept, and how to manage nuclear fuel waste. 16 refs., 1 tab., 3 figs

Used fuel management system architecture and interface analyses

Energy Technology Data Exchange (ETDEWEB)

Nutt, Mark [Argonne National Laboratory, Argonne, IL (United States); Howard, Robert; Busch, Ingrid [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Carter, Joe; Delley, Alexcia [Savannah River National Laboratory, Aiken, SC (United States); Hardin, Ernest; Kalinina, Elena [Sandia National Laboratories, Albuquerque NM (United States); Cotton, Thomas [Complex Systems LLC, Washington, DC (United States)

2013-07-01

between at-reactor used fuel management, consolidated storage facilities, and disposal facilities, along with the development of supporting logistics simulation tools, have been initiated to provide the U.S. Department of Energy (DOE) and other stakeholders with information regarding the various alternatives for managing used nuclear fuel (UNF) generated by the current fleet of light water reactors operating in the United States. An important UNF management system interface consideration is the need for ultimate disposal of UNF assemblies contained in waste packages that are sized to be compatible with different geologic media. Thermal analyses indicate that waste package sizes for the geologic media under consideration by the Used Fuel Disposition Campaign may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded and being loaded into the dry storage canisters currently in use. The implications of where and when the packaging or repackaging of commercial UNF will occur are key questions being addressed in this evaluation. The analysis demonstrated that thermal considerations will have a major impact on the operation of the system and that acceptance priority, rates, and facility start dates have significant system implications. (authors)

Used fuel management system architecture and interface analyses

International Nuclear Information System (INIS)

Nutt, Mark; Howard, Robert; Busch, Ingrid; Carter, Joe; Delley, Alexcia; Hardin, Ernest; Kalinina, Elena; Cotton, Thomas

2013-01-01

between at-reactor used fuel management, consolidated storage facilities, and disposal facilities, along with the development of supporting logistics simulation tools, have been initiated to provide the U.S. Department of Energy (DOE) and other stakeholders with information regarding the various alternatives for managing used nuclear fuel (UNF) generated by the current fleet of light water reactors operating in the United States. An important UNF management system interface consideration is the need for ultimate disposal of UNF assemblies contained in waste packages that are sized to be compatible with different geologic media. Thermal analyses indicate that waste package sizes for the geologic media under consideration by the Used Fuel Disposition Campaign may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded and being loaded into the dry storage canisters currently in use. The implications of where and when the packaging or repackaging of commercial UNF will occur are key questions being addressed in this evaluation. The analysis demonstrated that thermal considerations will have a major impact on the operation of the system and that acceptance priority, rates, and facility start dates have significant system implications. (authors)

Decontamination of stainless steel canisters that contain high-level waste

International Nuclear Information System (INIS)

Bray, L.A.

1987-01-01

At the West Valley Nuclear Services Company (WVNSC) in West Valley, New York, high-level radioactive waste (HLW) will be vitrified into a borosilicate glass form and poured into large, stainless steel canisters. During the filling process, volatile fission products, principally 137 Cs, condense on the exterior of the canisters. The smearable contamination that remains on the canisters after they are filled and partially cooled must be removed from the canisters' exterior surfaces prior to their storage and ultimate shipment to a US Department of Energy (DOE) repository for disposal. A simple and effective method was developed for decontamination of HLW canisters. This method of chemical decontamination is applicable to a wide variety of contaminated equipment found in the nuclear industry. The process employs a reduction-oxidation system [Ce(III)/Ce(IV)] in nitric acid solution to chemically mill the surface of stainless steel, similar to the electropolishing process, but without the need for an applied electrical current. Contaminated canisters are simply immersed in the solution at controlled temperature and Ce(IV) concentration levels

On the Impact of the Fuel Dissolution Rate Upon Near-Field Releases From Nuclear Waste Disposal

Directory of Open Access Journals (Sweden)

A Pereira

2016-09-01

Full Text Available Calculations of the impact of the dissolution of spent nuclear fuel on the release from a damaged canister in a KBS-3 repository are presented. The dissolution of the fuel matrix is a complex process and the dissolution rate is known to be one of the most important parameters in performance assessment models of the near-field of a geological repository. A variability study has been made to estimate the uncertainties associated with the process of fuel dissolution. The model considered in this work is a 3D model of a KBS-3 copper canister. The nuclide used in the calculations is Cs-135. Our results confirm that the fuel degradation rate is an important parameter, however there are considerable uncertainties associated with the data and the conceptual models. Consequently, in the interests of safety one should reduce, as far as possible, the uncertainties coupled to fuel degradation.

The role of the canister in a system for the final disposal of spent fuel or high-level waste

International Nuclear Information System (INIS)

Papp, T.

1986-01-01

A final repository for radioactive waste must isolate the toxic substances or distribute their release over time or space to avoid causing harmful concentrations of radionuclides in the biosphere. The Swedish research has focused on a repository 500 m down in crystalline rock where the geochemical environment can give canisters a service life of the order of a million years. These evaluations are discussed and the safety effect of the canister is compared with that of other barriers available in a repository system. Our conclusions are that a combined protection effect of natural and man-made barriers can be achieved that substantially exceeds what could reasonably be required by society. An actual repository design can then be based on an optimization of the cost to reach a level of accepted safety with due regard for the safety margins and redundancy necessary for achieving public confidence. (author)

Comparison of the waste management aspects of spent fuel disposal and reprocessing: post-disposal radiological impact

International Nuclear Information System (INIS)

Mobbs, S.F.; Harvey, M.P.; Martin, J.S.; Mayall, A.; Jones, M.E.

1991-01-01

A joint project involving contractors from France, Germany and the UK was set up by the Commission of the European Communities to assess the implications of two waste management options: the direct disposal of spent fuel and reprocessing of that fuel. This report describes the calculation of the radiological impact on the public of the management and disposal of the wastes associated with these two options. Six waste streams were considered: discharge of liquid reprocessing effluents, discharge of gaseous reprocessing effluents, disposal of low-level solid wastes arising from reprocessing, disposal of intermediate-level solid wastes arising from reprocessing, disposal of vitrified high-level reprocessing wastes, and direct disposal of spent fuel. The results of the calculations are in the form of maximum annual doses and risks to individual members of the public, and collective doses to four population groups, integrated over six time periods. These results were designed for input into a computer model developed by another contractor, Yard Ltd, which combines costs and impacts in a multi-attribute hierarchy to give an overall measure of the impact of a given option

Enhanced Thermal Management System for Spent Nuclear Fuel Dry Storage Canister with Hybrid Heat Pipes

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Bang, In Cheol

2016-01-01

Dry storage uses the gas or air as coolant within sealed canister with neutron shielding materials. Dry storage system for spent fuel is regarded as relatively safe and emits little radioactive waste for the storage, but it showed that the storage capacity and overall safety of dry cask needs to be enhanced for the dry storage cask for LWR in Korea. For safety enhancement of dry cask, previous studies of our group firstly suggested the passive cooling system with heat pipes for LWR spent fuel dry storage metal cask. As an extension, enhanced thermal management systems for the spent fuel dry storage cask for LWR was suggested with hybrid heat pipe concept, and their performances were analyzed in thermal-hydraulic viewpoint in this paper. In this paper, hybrid heat pipe concept for dry storage cask is suggested for thermal management to enhance safety margin. Although current design of dry cask satisfies the design criteria, it cannot be assured to have long term storage period and designed lifetime. Introducing hybrid heat pipe concept to dry storage cask designed without disrupting structural integrity, it can enhance the overall safety characteristics with adequate thermal management to reduce overall temperature as well as criticality control. To evaluate thermal performance of hybrid heat pipe according to its design, CFD simulation was conducted and previous and revised design of hybrid heat pipe was compared in terms of temperature inside canister

Disposal of spent nuclear fuel from NPP Krsko

International Nuclear Information System (INIS)

Mele, I.

2004-01-01

In order to get a clear view of the future liabilities of Slovenia and Croatia regarding the long term management of radioactive waste and spent nuclear fuel produced by the NPP Krsko, an estimation of disposal cost for low and intermediate level waste (LILW) as well as for spent nuclear fuel is needed. This cost estimation represents the basis for defining the target value for the financial resources to be accrued by the two national decommissioning and waste disposal funds, as determined in the agreement between Slovenia and Croatia on the ownership and exploitation of the NPP Krsko from March 2003, and for specifying their financial strategies. The one and only record of the NPP Krsko spent fuel disposal costs was made in the NPP Krsko Decommissioning Plan from 1996 [1]. As a result of incomplete input data, the above SF disposal cost estimate does not incorporate all cost elements. A new cost estimation was required in the process of preparation of the Joint Decommissioning and Waste Management Programme according to the provisions of the above mentioned agreement between Slovenia and Croatia. The basic presumptions and reference scenario for the disposal of spent nuclear fuel on which the cost estimation is based, as well as the applied methodology and results of cost estimation, are presented in this paper. Alternatives to the reference scenario and open questions which need to be resolved before the relevant final decision is taken, are also briefly discussed. (author)

Optimization of Deep Borehole Systems for HLW Disposal

International Nuclear Information System (INIS)

Driscoll, Michael; Baglietto, Emilio; Buongiorno, Jacopo; Lester, Richard; Brady, Patrick; Arnold, B. W.

2015-01-01

This is the final report on a project to update and improve the conceptual design of deep boreholes for high level nuclear waste disposal. The effort was concentrated on application to intact US legacy LWR fuel assemblies, but conducted in a way in which straightforward extension to other waste forms, host rock types and countries was preserved. The reference fuel design version consists of a vertical borehole drilled into granitic bedrock, with the uppermost kilometer serving as a caprock zone containing a diverse and redundant series of plugs. There follows a one to two kilometer waste canister emplacement zone having a hole diameter of approximately 40-50 cm. Individual holes are spaced 200-300 m apart to form a repository field. The choice of verticality and the use of a graphite based mud as filler between the waste canisters and the borehole wall liner was strongly influenced by the expectation that retrievability would continue to be emphasized in US and worldwide repository regulatory criteria. An advanced version was scoped out using zinc alloy cast in place to fill void space inside a disposal canister and its encapsulated fuel assembly. This excludes water and greatly improves both crush resistance and thermal conductivity. However the simpler option of using a sand fill was found adequate and is recommended for near-term use. Thermal-hydraulic modeling of the low permeability and porosity host rock and its small (@@@ 1%) saline water content showed that vertical convection induced by the waste's decay heat should not transport nuclides from the emplacement zone up to the biosphere atop the caprock. First order economic analysis indicated that borehole repositories should be cost-competitive with shallower mined repositories. It is concluded that proceeding with plans to drill a demonstration borehole to confirm expectations, and to carry out priority experiments, such as retention and replenishment of in-hole water is in order.

Optimization of Deep Borehole Systems for HLW Disposal

Energy Technology Data Exchange (ETDEWEB)

Driscoll, Michael [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Baglietto, Emilio [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Buongiorno, Jacopo [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Lester, Richard [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Brady, Patrick [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Arnold, B. W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-09-09

This is the final report on a project to update and improve the conceptual design of deep boreholes for high level nuclear waste disposal. The effort was concentrated on application to intact US legacy LWR fuel assemblies, but conducted in a way in which straightforward extension to other waste forms, host rock types and countries was preserved. The reference fuel design version consists of a vertical borehole drilled into granitic bedrock, with the uppermost kilometer serving as a caprock zone containing a diverse and redundant series of plugs. There follows a one to two kilometer waste canister emplacement zone having a hole diameter of approximately 40-50 cm. Individual holes are spaced 200-300 m apart to form a repository field. The choice of verticality and the use of a graphite based mud as filler between the waste canisters and the borehole wall liner was strongly influenced by the expectation that retrievability would continue to be emphasized in US and worldwide repository regulatory criteria. An advanced version was scoped out using zinc alloy cast in place to fill void space inside a disposal canister and its encapsulated fuel assembly. This excludes water and greatly improves both crush resistance and thermal conductivity. However the simpler option of using a sand fill was found adequate and is recommended for near-term use. Thermal-hydraulic modeling of the low permeability and porosity host rock and its small (≤ 1%) saline water content showed that vertical convection induced by the waste’s decay heat should not transport nuclides from the emplacement zone up to the biosphere atop the caprock. First order economic analysis indicated that borehole repositories should be cost-competitive with shallower mined repositories. It is concluded that proceeding with plans to drill a demonstration borehole to confirm expectations, and to carry out priority experiments, such as retention and replenishment of in-hole water is in order.

Drop tests of the Three Mile Island knockout canister

International Nuclear Information System (INIS)

Box, W.D.; Aaron, W.S.; Shappert, L.B.; Childress, P.C.; Quinn, G.J.; Smith, J.V.

1986-09-01

A type of Three Mile Island Unit 2 (TMI-2) defueling canister, called a ''knockout'' canister, was subjected to a series of drop tests at the Oak Ridge National Laboratory's Drop Test Facility. These tests were designed to confirm the structural integrity of internal fixed neutron poisons in support of a request for NRC licensing of this type of canister for the shipment of TMI-2 reactor fuel debris to the Idaho National Engineering Laboratory (INEL) for the Core Examination R and D Program. Work conducted at the Oak Ridge National Laboratory included (1) precise physical measurements of the internal poison rod configuration before assembly, (2) canister assembly and welding, (3) nondestructive examination (an initial hydrostatic pressure test and an x-ray profile of the internals before and after each drop test), (4) addition of a simulated fuel load, (5) instrumentation of the canister for each drop test, (6) fabrication of a cask simulation vessel with a developed and tested foam impact limiter, (7) use of refrigeration facilities to cool the canister to well below freezing prior to three of the drops, (8) recording the drop test with still, high-speed, and normal-speed photography, (9) recording the accelerometer measurements during impact, (10) disassembly and post-test examination with precise physical measurements, and (11) preparation of the final report

Geotechnical engineering for ocean waste disposal. An introduction

Science.gov (United States)

Lee, Homa J.; Demars, Kenneth R.; Chaney, Ronald C.; ,

1990-01-01

As members of multidisciplinary teams, geotechnical engineers apply quantitative knowledge about the behavior of earth materials toward designing systems for disposing of wastes in the oceans and monitoring waste disposal sites. In dredge material disposal, geotechnical engineers assist in selecting disposal equipment, predict stable characteristics of dredge mounds, design mound caps, and predict erodibility of the material. In canister disposal, geotechnical engineers assist in specifying canister configurations, predict penetration depths into the seafloor, and predict and monitor canister performance following emplacement. With sewage outfalls, geotechnical engineers design foundation and anchor elements, estimate scour potential around the outfalls, and determine the stability of deposits made up of discharged material. With landfills, geotechnical engineers evaluate the stability and erodibility of margins and estimate settlement and cracking of the landfill mass. Geotechnical engineers also consider the influence that pollutants have on the engineering behavior of marine sediment and the extent to which changes in behavior affect the performance of structures founded on the sediment. In each of these roles, careful application of geotechnical engineering principles can contribute toward more efficient and environmentally safe waste disposal operations.

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.)

Canister storage building design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

KOPELIC, S.D.

1999-02-25

This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

Safeguarding of spent fuel conditioning and disposal in geological repositories

International Nuclear Information System (INIS)

Forsstroem, H.; Richter, B.

1997-01-01

Disposal of spent nuclear fuel in geological formations, without reprocessing, is being considered in a number of States. Before disposal the fuel will be encapsulated in a tight and corrosion resistant container. The method chosen for disposal and the design of the repository will be determined by the geological conditions and the very strict requirements on long-term safety. From a safeguards perspective spent fuel disposal is a new issue. As the spent fuel still contains important amounts of material under safeguards and as it can not be considered practicably irrecoverable in the repository, the IAEA has been advised not to terminate safeguards, even after closure of the repository. This raises a number of new issues where there could be a potential conflict of interests between safety and safeguards demands, in particular in connection with the safety principle that burdens on future generations should be avoided. In this paper some of these issues are discussed based on the experience gained in Germany and Sweden about the design and future operation of encapsulation and disposal facilities. The most important issues are connected to the required level of safeguards for a closed repository, the differences in time scales for waste management and safeguards, the need for verification of the fissile content in the containers and the possibility of retrieving the fuel disposed of. (author)

Corrosion resistance of copper canister weld material

International Nuclear Information System (INIS)

Gubner, Rolf; Andersson, Urban

2007-03-01

The proposed design for a final repository for spent fuel and other long-lived residues is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast iron insert fitted inside a copper canister. SKB has since several years developed manufacturing processes for the canister components using a network of manufacturers. For the encapsulation process SKB has built the Canister Laboratory to demonstrate and develop the encapsulation technique in full scale. The critical part of the encapsulation of spent fuel is the sealing of the canister which is done by welding the copper lid to the cylindrical part of the canister. Two welding techniques have been developed in parallel, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). During the past two decades, SKB has developed the technology EBW at The Welding Institute (TWI) in Cambridge, UK. The development work at the Canister Laboratory began in 1999. In electron beam welding, a gun is used to generate the electron beam which is aimed at the joint. The beam heats up the material to the melting point allowing a fusion weld to be formed. The gun was developed by TWI and has a unique design for use at reduced pressure. The system has gone through a number of improvements under the last couple of years including implementation of a beam oscillation system. However, during fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities at the new grain boundaries that might pose adverse effects on the corrosion resistance of welds. As a new method for joining, SKB has been developing friction stir welding (FSW) for sealing copper canisters for spent nuclear fuel in cooperation with TWI since 1997. FSW was invented in 1991 at TWI and is a thermo

Corrosion resistance of copper canister weld material

Energy Technology Data Exchange (ETDEWEB)

Gubner, Rolf; Andersson, Urban [Corrosion and Metals Research Institute, Sto ckholm (Sweden)

2007-03-15

The proposed design for a final repository for spent fuel and other long-lived residues is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast iron insert fitted inside a copper canister. SKB has since several years developed manufacturing processes for the canister components using a network of manufacturers. For the encapsulation process SKB has built the Canister Laboratory to demonstrate and develop the encapsulation technique in full scale. The critical part of the encapsulation of spent fuel is the sealing of the canister which is done by welding the copper lid to the cylindrical part of the canister. Two welding techniques have been developed in parallel, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). During the past two decades, SKB has developed the technology EBW at The Welding Institute (TWI) in Cambridge, UK. The development work at the Canister Laboratory began in 1999. In electron beam welding, a gun is used to generate the electron beam which is aimed at the joint. The beam heats up the material to the melting point allowing a fusion weld to be formed. The gun was developed by TWI and has a unique design for use at reduced pressure. The system has gone through a number of improvements under the last couple of years including implementation of a beam oscillation system. However, during fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities at the new grain boundaries that might pose adverse effects on the corrosion resistance of welds. As a new method for joining, SKB has been developing friction stir welding (FSW) for sealing copper canisters for spent nuclear fuel in cooperation with TWI since 1997. FSW was invented in 1991 at TWI and is a thermo

Cost analysis for application of solidified waste fission product canisters in U.S. Army steam plants

International Nuclear Information System (INIS)

Sande, W.E.; Bjorklund, W.J.; Brooks, N.A.

1977-04-01

The main objectives of the present study are to design steam plants using projected waste fission product canister characteristics, to analyze the overall impact and cost/benefit to the nuclear fuel cycle associated with these plants, and to develop plans for this application if the cost analysis so warrants it. The construction and operation of a steam plant fueled with waste fission product canisters would require the involvement and cooperation of various government agencies and private industry; thus the philosophies of these groups were studied. These philosophies are discussed, followed by a forecast of canister supply, canister characteristics, and strategies for Army canister use. Another section describes the safety and licensing of these steam plants since this affects design and capital costs. The discussion of steam plant design includes boiler concepts, boiler heat transfer, canister temperature distributions, steam plant size, and steam plant operation. Also, canister transportation is discussed since this influences operating costs. Details of economics of Army steam plants are provided including steam plant capital costs, operating costs, fuel reprocessor savings due to Army canister storage, and overall economics. Recommendations are made in the final section

Geological Disposal of Nuclear Waste: Investigating the Thermo-Hygro-Mechanical-Chemical (THMC) Coupled Processes at the Waste Canister- Bentonite Barrier Interface

Science.gov (United States)

Davies, C. W.; Davie, D. C.; Charles, D. A.

2015-12-01

Geological disposal of nuclear waste is being increasingly considered to deal with the growing volume of waste resulting from the nuclear legacy of numerous nations. Within the UK there is 650,000 cubic meters of waste safely stored and managed in near-surface interim facilities but with no conclusive permanent disposal route. A Geological Disposal Facility with incorporated Engineered Barrier Systems are currently being considered as a permanent waste management solution (Fig.1). This research focuses on the EBS bentonite buffer/waste canister interface, and experimentally replicates key environmental phases that would occur after canister emplacement. This progresses understanding of the temporal evolution of the EBS and the associated impact on its engineering, mineralogical and physicochemical state and considers any consequences for the EBS safety functions of containment and isolation. Correlation of engineering properties to the physicochemical state is the focus of this research. Changes to geotechnical properties such as Atterberg limits, swelling pressure and swelling kinetics are measured after laboratory exposure to THMC variables from interface and batch experiments. Factors affecting the barrier, post closure, include corrosion product interaction, precipitation of silica, near-field chemical environment, groundwater salinity and temperature. Results show that increasing groundwater salinity has a direct impact on the buffer, reducing swelling capacity and plasticity index by up to 80%. Similarly, thermal loading reduces swelling capacity by 23% and plasticity index by 5%. Bentonite/steel interaction studies show corrosion precipitates diffusing into compacted bentonite up to 3mm from the interface over a 4 month exposure (increasing with temperature), with reduction in swelling capacity in the affected zone, probably due to the development of poorly crystalline iron oxides. These results indicate that groundwater conditions, temperature and corrosion

Used Fuel Management System Interface Analyses - 13578

Energy Technology Data Exchange (ETDEWEB)

Howard, Robert; Busch, Ingrid [Oak Ridge National Laboratory, P.O. Box 2008, Bldg. 5700, MS-6170, Oak Ridge, TN 37831 (United States); Nutt, Mark; Morris, Edgar; Puig, Francesc [Argonne National Laboratory (United States); Carter, Joe; Delley, Alexcia; Rodwell, Phillip [Savannah River National Laboratory (United States); Hardin, Ernest; Kalinina, Elena [Sandia National Laboratories (United States); Clark, Robert [U.S. Department of Energy (United States); Cotton, Thomas [Complex Systems Group (United States)

2013-07-01

Preliminary system-level analyses of the interfaces between at-reactor used fuel management, consolidated storage facilities, and disposal facilities, along with the development of supporting logistics simulation tools, have been initiated to provide the U.S. Department of Energy (DOE) and other stakeholders with information regarding the various alternatives for managing used nuclear fuel (UNF) generated by the current fleet of light water reactors operating in the United States. An important UNF management system interface consideration is the need for ultimate disposal of UNF assemblies contained in waste packages that are sized to be compatible with different geologic media. Thermal analyses indicate that waste package sizes for the geologic media under consideration by the Used Fuel Disposition Campaign may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded and being loaded into the dry storage canisters currently in use. The implications of where and when the packaging or repackaging of commercial UNF will occur are key questions being addressed in this evaluation. The analysis demonstrated that thermal considerations will have a major impact on the operation of the system and that acceptance priority, rates, and facility start dates have significant system implications. (authors)

Clay Generic Disposal System Model - Sensitivity Analysis for 32 PWR Assembly Canisters (+2 associated model files).

Energy Technology Data Exchange (ETDEWEB)

Morris, Edgar [Argonne National Lab. (ANL), Argonne, IL (United States)

2014-10-01

The Used Fuel Disposition Campaign (UFDC), as part of the DOE Office of Nuclear Energy’s (DOE-NE) Fuel Cycle Technology program (FCT) is investigating the disposal of high level radioactive waste (HLW) and spent nuclear fuela (SNF) in a variety of geologic media. The feasibility of disposing SNF and HLW in clay media has been investigated and has been shown to be promising [Ref. 1]. In addition the disposal of these wastes in clay media is being investigated in Belgium, France, and Switzerland. Thus, Argillaceous media is one of the environments being considered by UFDC. As identified by researchers at Sandia National Laboratory, potentially suitable formations that may exist in the U.S. include mudstone, clay, shale, and argillite formations [Ref. 1]. These formations encompass a broad range of material properties. In this report, reference to clay media is intended to cover the full range of material properties. This report presents the status of the development of a simulation model for evaluating the performance of generic clay media. The clay Generic Disposal System Model (GDSM) repository performance simulation tool has been developed with the flexibility to evaluate not only different properties, but different waste streams/forms and different repository designs and engineered barrier configurations/ materials that could be used to dispose of these wastes.

SCFR Fuel Cycles and Their Impact on the Performance of High-Level Waste Disposal

Energy Technology Data Exchange (ETDEWEB)

Kawasaki, Daisuke; Nogi, Naoyuki; Saito, Takumi [Department of Nuclear Engineering and Management, Graduate School of Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo, 113-8656 (Japan); Nagasaki, Shinya [Nuclear Professional School, Graduate School of Engineering, The University of Tokyo, 2-22 Shirakata Shirane, Tokai, Ibaraki, 319-1188 (Japan)

2009-06-15

The concept of supercritical-pressure light water cooled fast reactor (SCFR) is developed and studied at the University of Tokyo. The impact of disposal of the waste generated in a fuel cycle with SCFR is also investigated in the project. It is of great interest how a fuel cycle with SCFR compares to the other fuel cycles from the back-end view point. With its various neutron spectrum, SCFR may be used to transmute both actinides and fission products. The objective of the present study is to evaluate and compare multiple fuel cycle designs in order to investigate the effects of SCFR and its transmutation capability upon the back-end risks. Three designs of fuel cycle are considered for evaluation in the present study. First, a simple fuel cycle with PWR and recycling is considered. The spent fuel from the PWR is reprocessed to recover uranium and plutonium, and the rest of the radioactive nuclides are vitrified and disposed of in a geologic repository. In the second design, the recovered uranium and plutonium in the reprocessing of PWR spent fuel is fabricated into a MOX fuel and irradiated in SCFR. The spent fuel from the SCFR is reprocessed to recover uranium and plutonium. In the third design, actinide elements are also separated from the PWR spent fuel and is loaded as the blanket fuel in SCFR core together with the MOX fuel fabricated from the recovered uranium and plutonium. In the same way as in the second design, the spent fuel from the SCFR is reprocessed to recover uranium and plutonium. In the second and the third designs, there are two streams of highly radioactive waste; one from the reprocessing (separation process) of the PWR spent fuel, and the other from the reprocessing of the SCFR spent fuel. Numerical codes Origen2.1 and SWAT is used for fuel irradiation calculation. The performance of the high-level radioactive waste repository is evaluated for each design of fuel cycle. It is assumed that the repository is located in a water-saturated geologic

Waste disposal options report. Volume 2

International Nuclear Information System (INIS)

Russell, N.E.; McDonald, T.G.; Banaee, J.; Barnes, C.M.; Fish, L.W.; Losinski, S.J.; Peterson, H.K.; Sterbentz, J.W.; Wenzel, D.R.

1998-02-01

Volume 2 contains the following topical sections: estimates of feed and waste volumes, compositions, and properties; evaluation of radionuclide inventory for Zr calcine; evaluation of radionuclide inventory for Al calcine; determination of k eff for high level waste canisters in various configurations; review of ceramic silicone foam for radioactive waste disposal; epoxides for low-level radioactive waste disposal; evaluation of several neutralization cases in processing calcine and sodium-bearing waste; background information for EFEs, dose rates, watts/canister, and PE-curies; waste disposal options assumptions; update of radiation field definition and thermal generation rates for calcine process packages of various geometries-HKP-26-97; and standard criteria of candidate repositories and environmental regulations for the treatment and disposal of ICPP radioactive mixed wastes

Waste disposal options report. Volume 2

Energy Technology Data Exchange (ETDEWEB)

Russell, N.E.; McDonald, T.G.; Banaee, J.; Barnes, C.M.; Fish, L.W.; Losinski, S.J.; Peterson, H.K.; Sterbentz, J.W.; Wenzel, D.R.

1998-02-01

Volume 2 contains the following topical sections: estimates of feed and waste volumes, compositions, and properties; evaluation of radionuclide inventory for Zr calcine; evaluation of radionuclide inventory for Al calcine; determination of k{sub eff} for high level waste canisters in various configurations; review of ceramic silicone foam for radioactive waste disposal; epoxides for low-level radioactive waste disposal; evaluation of several neutralization cases in processing calcine and sodium-bearing waste; background information for EFEs, dose rates, watts/canister, and PE-curies; waste disposal options assumptions; update of radiation field definition and thermal generation rates for calcine process packages of various geometries-HKP-26-97; and standard criteria of candidate repositories and environmental regulations for the treatment and disposal of ICPP radioactive mixed wastes.

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

CROWE, R.D.; PIEPHO, M.G.

2000-01-01

This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

CROWE, R.D.

1999-01-01

This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

Fuel removal, transport, and storage

International Nuclear Information System (INIS)

Reno, H.W.

1986-01-01

The March 1979 accident at Unit 2 of the Three Mile Island Nuclear Power Station (TMI-2) which damaged the core of the reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing the core debris from the reactor, packaging it into canisters, loading canisters into a rail cask, and transporting the debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights how some challenges were resolved, including lessons learned and benefits derived therefrom. Key to some success at TMI was designing, testing, fabricating, and licensing two rail casks, which each provide double containment of the damaged fuel. 10 refs., 12 figs

Safety case for the disposal of spent nuclear fuel at Olkiluoto. Features, events and processes 2012

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-12-15

Features, Events and Processes sits within Posiva Oy's Safety Case 'TURVA-2012' portfolio and has the objective of presenting the main features, events and processes (FEPs) that are considered to be potentially significant for the long-term safety of the planned KBS-3V repository for spent nuclear fuel at Olkiluoto. The primary purpose of this report is to support Performance Assessment, Formulation of Radionuclide Release Scenarios, Assessment of the Radionuclide Release Scenarios for the Repository System and Biosphere Assessment by ensuring that the scenarios are comprehensive and take account of all significant FEPs. The main FEPs potentially affecting the disposal system are described for each relevant subsystem component or barrier (i.e. the spent nuclear fuel, the canister, the buffer and tunnel backfill, the auxiliary components, the geosphere and the surface environment). In addition, a small number of external FEPs that may potentially influence the evolution of the disposal system are described. The conceptual understanding and operation of each FEP is described, together with the main features (variables) of the disposal system that may affect its occurrence or significance. Olkiluoto-specific issues are considered when relevant. The main uncertainties (conceptual and parameter/data) associated with each FEP that may affect understanding are also documented. Indicative parameter values are provided, in some cases, to illustrate the magnitude or rate of a process, but it is not the intention of this report to provide the complete set of numerical values that are used in the quantitative safety assessment calculations. Many of the FEPs are interdependent and, therefore, the descriptions also identify the most important direct couplings between the FEPs. This information is used in the formulation of scenarios to ensure the conceptual models and calculational cases are both comprehensive and representative. (orig.)

Safety case for the disposal of spent nuclear fuel at Olkiluoto. Features, events and processes 2012

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-12-15

Features, Events and Processes sits within Posiva Oy's Safety Case 'TURVA-2012' portfolio and has the objective of presenting the main features, events and processes (FEPs) that are considered to be potentially significant for the long-term safety of the planned KBS-3V repository for spent nuclear fuel at Olkiluoto. The primary purpose of this report is to support Performance Assessment, Formulation of Radionuclide Release Scenarios, Assessment of the Radionuclide Release Scenarios for the Repository System and Biosphere Assessment by ensuring that the scenarios are comprehensive and take account of all significant FEPs. The main FEPs potentially affecting the disposal system are described for each relevant subsystem component or barrier (i.e. the spent nuclear fuel, the canister, the buffer and tunnel backfill, the auxiliary components, the geosphere and the surface environment). In addition, a small number of external FEPs that may potentially influence the evolution of the disposal system are described. The conceptual understanding and operation of each FEP is described, together with the main features (variables) of the disposal system that may affect its occurrence or significance. Olkiluoto-specific issues are considered when relevant. The main uncertainties (conceptual and parameter/data) associated with each FEP that may affect understanding are also documented. Indicative parameter values are provided, in some cases, to illustrate the magnitude or rate of a process, but it is not the intention of this report to provide the complete set of numerical values that are used in the quantitative safety assessment calculations. Many of the FEPs are interdependent and, therefore, the descriptions also identify the most important direct couplings between the FEPs. This information is used in the formulation of scenarios to ensure the conceptual models and calculational cases are both comprehensive and representative. (orig.)

Safety case for the disposal of spent nuclear fuel at Olkiluoto. Features, events and processes 2012

International Nuclear Information System (INIS)

2012-12-01

Features, Events and Processes sits within Posiva Oy's Safety Case 'TURVA-2012' portfolio and has the objective of presenting the main features, events and processes (FEPs) that are considered to be potentially significant for the long-term safety of the planned KBS-3V repository for spent nuclear fuel at Olkiluoto. The primary purpose of this report is to support Performance Assessment, Formulation of Radionuclide Release Scenarios, Assessment of the Radionuclide Release Scenarios for the Repository System and Biosphere Assessment by ensuring that the scenarios are comprehensive and take account of all significant FEPs. The main FEPs potentially affecting the disposal system are described for each relevant subsystem component or barrier (i.e. the spent nuclear fuel, the canister, the buffer and tunnel backfill, the auxiliary components, the geosphere and the surface environment). In addition, a small number of external FEPs that may potentially influence the evolution of the disposal system are described. The conceptual understanding and operation of each FEP is described, together with the main features (variables) of the disposal system that may affect its occurrence or significance. Olkiluoto-specific issues are considered when relevant. The main uncertainties (conceptual and parameter/data) associated with each FEP that may affect understanding are also documented. Indicative parameter values are provided, in some cases, to illustrate the magnitude or rate of a process, but it is not the intention of this report to provide the complete set of numerical values that are used in the quantitative safety assessment calculations. Many of the FEPs are interdependent and, therefore, the descriptions also identify the most important direct couplings between the FEPs. This information is used in the formulation of scenarios to ensure the conceptual models and calculational cases are both comprehensive and representative. (orig.)

CFD Analysis on the Passive Heat Removal by Helium and Air in the Canister of Spent Fuel Dry Storage System

Energy Technology Data Exchange (ETDEWEB)

Shin, Do Young; Jeong, Ui Ju; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

2016-05-15

In the current commercial design, the canister of the dry storage system is mainly backfilled with helium gas. Helium gas shows very conductive behavior due to high thermal conductivity and small density change with temperature. However, other gases such as air, argon, or nitrogen are expected to show effective convective behavior. Thus these are also considered as candidates for the backfill gas to provide effective coolability. In this study, to compare the dominant cooling mechanism and effectiveness of cooling between helium gas and air, a computational fluid dynamics (CFD) analysis for the canister of spent fuel dry storage system with backfill gas of helium and air is carried out. In this study, CFD simulations for the helium and air backfilled gas for dry storage system canister were carried out using ANSYS FLUENT code. For the comparison work, two backfilled fluids were modeled with same initial and boundary conditions. The observed major difference can be summarized as follows. - The simulation results showed the difference in dominant heat removal mechanism. Conduction for helium, and convection for air considering Reynolds number distribution. - The temperature gradient inside the fuel assembly showed that in case of air, more effective heat mixing occurred compared to helium.

Competitive effect of metallic canister and clay barrier on the sorption of Eu3+ under subcritical conditions

International Nuclear Information System (INIS)

El Mrabet, Said; Castro, Miguel A.; Orta, M. Mar; Pazos, M. Carolina; Alba, Maria D.; Astudillo, Julio; Rueda, Silvia; Hurtado, Santiago; Villa, Mara

2012-01-01

Document available in extended abstract form only. The disposal of high level radioactive wastes (HLW) such as spent fuel or reprocessing waste resulting from the operation and dismantling of nuclear reactors is one of the most problems facing the worlds because of its long half life and radionuclide migration to the biosphere. For long term performance assessment of radioactive waste disposal, knowledge concerning radionuclide retention processes on materials composing the engineered barrier (clay and container waste) is required. Steel waste containers and bentonite have been proposed as candidate materials for overpack and buffer respectively in most of the proposed repositories designs for nuclear waste disposal. This contribution aims to study the competitiveness of the bentonite and the metallic canister in the retention process of some kinds of radioactive waste such as 152 Eu. The europium was chosen because it is a toxic metal and usually taken as a simulator of the trivalent high level radioactive waste. In order to elucidate the mechanisms involved in the retention processes of europium by both bentonite and metallic canister, a cylindrical steel mini-reactor was designed and prepared from the same material as the steel reactor AISI-316L. The bentonite was then introduced and compacted within the mini-reactor forming a set mini-reactor- bentonite. The system mini-reactor -bentonite was then subjected to hydrothermal treatments at 300 deg. C for 4.5 days. The morphology and chemical composition of both steel and bentonite were analyzed by XRD and SEM. SEM and XRD results revealed that both the bentonite and the metallic canister were involved in sorption mechanism of europium by the formation of insoluble phases of europium silicates originated from the mixed solution of bentonite, Eu 3+ and canister. The pH-Redox potential (Eh) indicated that the interlayer cations of bentonite were replaced by Eu 3+ with higher acidity and Eh which means that the active

Nuclear fuel waste disposal

International Nuclear Information System (INIS)

1982-01-01

This film for a general audience deals with nuclear fuel waste management in Canada, where research is concentrating on land based geologic disposal of wastes rather than on reprocessing of fuel. The waste management programme is based on cooperation of the AECL, various universities and Ontario Hydro. Findings of research institutes in other countries are taken into account as well. The long-term effects of buried radioactive wastes on humans (ground water, food chain etc.) are carefully studied with the help of computer models. Animated sequences illustrate the behaviour of radionuclides and explain the idea of a multiple barrier system to minimize the danger of radiation hazards

DOE requests waiver on double containment for HLW canisters

International Nuclear Information System (INIS)

Lobsenz, G.

1994-01-01

The Energy Department has asked the Nuclear Regulatory Commission to waive double containment requirements for vitrified high-level radioactive waste canisters, saying the additional protection is not necessary and too costly. NRC said it had received a petition from DOE contending that the vitrified waste canisters were durable enough without double containment to prevent any potential plutonium release during handling and shipping. DOE said testing had shown that the vitrified waste canisters were similar - even superior - in durability to spent reactor fuel shipments, which NRC specifically exempted from the double containment requirement

The deep disposal repository - an alternative general arrangement

International Nuclear Information System (INIS)

Wolofsky, L.

1996-01-01

For the disposal of nuclear fuel waste in Canada the current proposal requires a repository at or below 500 m depth in Precambrian plutonic igneous rock in Ontario, consisting of numerous horizontal parallel tunnels arranged in one or more horizontal planes i.e. a room-and-pillar arrangement. The pillars are three times as wide as the tunnels, giving as extraction ratio of 25%. The tunnels are tentatively designed to be excavated by drill and blast, with a flat-arched roof, vertical side walls and a flat invert. The fuel containing canisters are to be surrounded by a low-permeability bentonite-based buffer in vertical holes drilled for the purpose in the invert

Krypton-85 disposal program. Semiannual report, August 15, 1977--March 31, 1978

International Nuclear Information System (INIS)

Klett, R.D.

1979-02-01

The first 7.5 months of the Krypton-85 disposal program are summarized. Included are task definitions and initial progress in geologic disposal system studies, SURF compatibility, augmented heat dissipation, material qualification, exterior canister compatibility, ceramic liners for canisters, and geologic transport. Feasibility studies indicate that Kr-85 can be disposed of at SURF facility or in near-surface geologic repositories

Debris Removal Project K West Canister Cleaning System Performance Specification

International Nuclear Information System (INIS)

FARWICK, C.C.

1999-01-01

Approximately 2,300 metric tons Spent Nuclear Fuel (SNF) are currently stored within two water filled pools, the 105 K East (KE) fuel storage basin and the 105 K West (KW) fuel storage basin, at the U.S. Department of Energy, Richland Operations Office (RL). The SNF Project is responsible for operation of the K Basins and for the materials within them. A subproject to the SNF Project is the Debris Removal Subproject, which is responsible for removal of empty canisters and lids from the basins. Design criteria for a Canister Cleaning System to be installed in the KW Basin. This documents the requirements for design and installation of the system

The Canadian program for management of spent fuel and high level wastes

International Nuclear Information System (INIS)

Barnes, R.W.; Mayman, S.A.

A brief history and description of the nuclear power program in Canada is given. Schedules and programs are described for storing spent fuel in station fuel bays, centralized water pool storage facilities, concrete canisters, convection vaults, and rock or salt formations. High-level wastes will be retrievable initially, therefore the focus is on storage in mined cavities. The methods developed for high-level waste storage/disposal will ideally be flexible enough to accommodate spent fuel. (E.C.B.)

Treatment and disposal of radioactive wastes from nuclear power plants. Program for encapsulation, deep geologic deposition and research, development and demonstration; Kaernkraftavfallets behandling och slutfoervaring. Program foer inkapsling, geologisk djupfoervaring samt forskning, utveckling och demonstration

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

Programs for RD and D concerning disposal of radioactive waste are presented. Main topics include: Design, testing and manufacture of canisters for the spent fuels; Design of equipment for deposition of waste canisters; Material and process for backfilling rock caverns; Evaluation of accuracy and validation of methods for safety analyses; Development of methods for defining scenarios for the safety analyses. 471 refs, 67 figs, 21 tabs.

Value Engineering Study for Closing Waste Packages Containing TAD Canisters

Energy Technology Data Exchange (ETDEWEB)

Colleen Shelton-Davis

2005-11-01

The Office of Civilian Radioactive Waste Management announced their intention to have the commercial utilities package spent nuclear fuel in shielded, transportable, ageable, and disposable containers prior to shipment to the Yucca Mountain repository. This will change the conditions used as a basis for the design of the waste package closure system. The environment is now expected to be a low radiation, low contamination area. A value engineering study was completed to evaluate possible modifications to the existing closure system using the revised requirements. Four alternatives were identified and evaluated against a set of weighted criteria. The alternatives are (1) a radiation-hardened, remote automated system (the current baseline design); (2) a nonradiation-hardened, remote automated system (with personnel intervention if necessary); (3) a nonradiation-hardened, semi-automated system with personnel access for routine manual operations; and (4) a nonradiation-hardened, fully manual system with full-time personnel access. Based on the study, the recommended design is Alternative 2, a nonradiation-hardened, remote automated system. It is less expensive and less complex than the current baseline system, because nonradiation-hardened equipment can be used and some contamination control equipment is no longer needed. In addition, the inclusion of remote automation ensures throughput requirements are met, provides a more reliable process, and provides greater protection for employees from industrial accidents and radiation exposure than the semi-automated or manual systems. Other items addressed during the value engineering study as requested by OCRWM include a comparison to industry canister closure systems and corresponding lessons learned; consideration of closing a transportable, ageable, and disposable canister; and an estimate of the time required to perform a demonstration of the recommended closure system.

Value Engineering Study for Closing Waste Packages Containing TAD Canisters

International Nuclear Information System (INIS)

Colleen Shelton-Davis

2005-01-01

The Office of Civilian Radioactive Waste Management announced their intention to have the commercial utilities package spent nuclear fuel in shielded, transportable, ageable, and disposable containers prior to shipment to the Yucca Mountain repository. This will change the conditions used as a basis for the design of the waste package closure system. The environment is now expected to be a low radiation, low contamination area. A value engineering study was completed to evaluate possible modifications to the existing closure system using the revised requirements. Four alternatives were identified and evaluated against a set of weighted criteria. The alternatives are (1) a radiation-hardened, remote automated system (the current baseline design); (2) a nonradiation-hardened, remote automated system (with personnel intervention if necessary); (3) a nonradiation-hardened, semi-automated system with personnel access for routine manual operations; and (4) a nonradiation-hardened, fully manual system with full-time personnel access. Based on the study, the recommended design is Alternative 2, a nonradiation-hardened, remote automated system. It is less expensive and less complex than the current baseline system, because nonradiation-hardened equipment can be used and some contamination control equipment is no longer needed. In addition, the inclusion of remote automation ensures throughput requirements are met, provides a more reliable process, and provides greater protection for employees from industrial accidents and radiation exposure than the semi-automated or manual systems. Other items addressed during the value engineering study as requested by OCRWM include a comparison to industry canister closure systems and corresponding lessons learned; consideration of closing a transportable, ageable, and disposable canister; and an estimate of the time required to perform a demonstration of the recommended closure system

Report on the disposal of radioactive wastes and spent fuel elements from Baden-Wuerttemberg

International Nuclear Information System (INIS)

2017-04-01

The report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg covers the following issues: legal framework for the nuclear disposal; producer of spent fuels and radioactive wastes in Baden- Report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg; low- and medium-level radioactive wastes (non heat generating radioactive wastes); spent fuels and radioactive wastes from waste processing (heat generating radioactive wastes); final disposal.

Iron-nickel alloys as canister material for radioactive waste disposal in underground repositories

International Nuclear Information System (INIS)

Apps, J.A.

1982-01-01

Canisters containing high-level radioactive waste must retain their integrity in an underground waste repository for at least one thousand years after burial (Nuclear Regulatory Commission, 1981). Since no direct means of verifying canister integrity is plausible over such a long period, indirect methods must be chosen. A persuasive approach is to examine the natural environment and find a suitable material which is thermodynamically compatible with the host rock under the environmental conditions with the host rock under the environmental conditions expected in a waste repository. Several candidates have been proposed, among them being iron-nickel alloys that are known to occur naturally in altered ultramafic rocks. The following review of stability relations among iron-nickel alloys below 350 0 C is the initial phase of a more detailed evaluation of these alloys as suitable canister materials

Program for responsible and safe disposal of spent fuel elements and radioactive wastes (National disposal program)

International Nuclear Information System (INIS)

2015-01-01

The contribution covers the following topics: fundamentals of the disposal policy; amount of radioactive wastes and prognosis; disposal of radioactive wastes - spent fuel elements and wastes from waste processing, radioactive wastes with low heat production; legal framework of the nuclear waste disposal in Germany; public participation, cost and financing.

HLW Canister and Can-In-Canister Drop Calculation

International Nuclear Information System (INIS)

H. Marr

1999-01-01

The purpose of this calculation is to evaluate the structural response of the standard high-level waste (HLW) canister and the HLW canister containing the cans of immobilized plutonium (''can-in-canister'' throughout this document) to the drop event during the handling operation. The objective of the calculation is to provide the structure parameter information to support the canister design and the waste handling facility design. Finite element solution is performed using the commercially available ANSYS Version (V) 5.4 finite element code. Two-dimensional (2-D) axisymmetric and three-dimensional (3-D) finite element representations for the standard HLW canister and the can-in-canister are developed and analyzed using the dynamic solver

The analysis of the RA reactor irradiated fuel cooling in the spent fuel pool; Analiza hladjenja ozracenog goriva u bazenu za odlaganje reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Vrhovac, M; Afgan, N; Spasojevic, D; Jovic, V [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1985-07-01

According to the RA reactor exploitation plan the great quantity of the irradiated spent fuel will be disposed in the reactor spent fuel pool after each reactor campaign which will including the present spent fuel inventory increase the residual power level in the pool and will soon cause the pool capacity shortage. To enable the analysis of the irradiated fuel cooling the pool and characteristic spent fuel canister temperature distribution at the residual power maximum was done. The results obtained under the various spent fuel cooling conditions in the pit indicate the normal spent fuel thermal load even in the most inconvenient cooling conditions. (author)

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

1999-09-09

This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

Measurements of Fundamental Fluid Physics of SNF Storage Canisters

Energy Technology Data Exchange (ETDEWEB)

Condie, Keith Glenn; Mc Creery, Glenn Ernest; McEligot, Donald Marinus

2001-09-01

With the University of Idaho, Ohio State University and Clarksean Associates, this research program has the long-term goal to develop reliable predictive techniques for the energy, mass and momentum transfer plus chemical reactions in drying / passivation (surface oxidation) operations in the transfer and storage of spent nuclear fuel (SNF) from wet to dry storage. Such techniques are needed to assist in design of future transfer and storage systems, prediction of the performance of existing and proposed systems and safety (re)evaluation of systems as necessary at later dates. Many fuel element geometries and configurations are accommodated in the storage of spent nuclear fuel. Consequently, there is no one generic fuel element / assembly, storage basket or canister and, therefore, no single generic fuel storage configuration. One can, however, identify generic flow phenomena or processes which may be present during drying or passivation in SNF canisters. The objective of the INEEL tasks was to obtain fundamental measurements of these flow processes in appropriate parameter ranges.

Test plan for K Basin Sludge Canister and Floor Sampling Device

International Nuclear Information System (INIS)

Meling, T.A.

1995-01-01

This document provides the test plan and procedure forms for conducting the functional and operational acceptance testing of the K Basin Sludge Canister and Floor Sampling Device(s). These samplers samples sludge off the floor of the 100K Basins and out of 100K fuel storage canisters

Gas liquid sampling for closed canisters in KW Basin - test plan

International Nuclear Information System (INIS)

Pitkoff, C.C.

1995-01-01

Test procedures for the gas/liquid sampler. Characterization of the Spent Nuclear Fuel, SNF, sealed in canisters at KW-Basin is needed to determine the state of storing SNF wet. Samples of the liquid and the gas in the closed canisters will be taken to gain characterization information. Sampling equipment has been designed to retrieve gas and liquid from the closed canisters in KW basin. This plan is written to outline the test requirements for this developmental sampling equipment

Cost Comparison for the Transfer of Select Calcined Waste Canisters to the Monitored Geologic Repository at Yucca Mountain, NV

International Nuclear Information System (INIS)

Michael B. Heiser; Clark B. Millet

2005-01-01

This report performs a life-cycle cost comparison of three proposed canister designs for the shipment and disposition of Idaho National Laboratory high-level calcined waste currently in storage at the Idaho Nuclear Technology and Engineering Center to the proposed national monitored geologic repository at Yucca Mountain, Nevada. Concept A (2 x 10-ft) and Concept B (2 x 15-ft) canisters are comparable in design, but they differ in size and waste loading options and vary proportionally in weight. The Concept C (5.5 x 17.5-ft) canister (also called the ''super canister''), while similar in design to the other canisters, is considerably larger and heavier than Concept A and B canisters and has a greater wall thickness. This report includes estimating the unique life-cycle costs for the three canister designs. Unique life-cycle costs include elements such as canister purchase and filling at the Idaho Nuclear Technology and Engineering Center, cask preparation and roundtrip consignment costs, final disposition in the monitored geologic repository (including canister off-loading and placement in the final waste disposal package for disposition), and cask purchase. Packaging of the calcine ''as-is'' would save $2.9 to $3.9 billion over direct vitrification disposal in the proposed national monitored geologic repository at Yucca Mountain, Nevada. Using the larger Concept C canisters would use 0.75 mi less of tunnel space, cost $1.3 billion less than 10-ft canisters of Concept A, and would be complete in 6.2 years

Review of geoscientific data of relevance to disposal of spent nuclear fuel in deep boreholes in crystalline rock

International Nuclear Information System (INIS)

Marsic, Nico; Grundfelt, Bertil

2013-09-01

In this report a compilation of recent geoscientific data of relevance to disposal of spent nuclear fuel in deep boreholes in Sweden is presented. The goal of the study has been limited to identifying and briefly describing such geoscientific information of relevance to disposal in deep boreholes that was not available at the time when previous compilations were made. Hence, the study is not to be regarded as a general up-date of new geoscientific information. Disposal of spent nuclear fuel in deep boreholes has been studied in Sweden since the second half of the 1980s. The currently studied concept has been proposed by Sandia National Laboratories in the USA. In this concept the spent fuel elements are encapsulated in cylindrical steel canisters that are joined together in strings of 40 canisters and lowered into five kilometres deep boreholes. Ten such strings are stacked between three and five kilometres depth separated from each other by concrete plugs. The study started with a review of boreholes that have been reported after the previous reviews that were published in 1998 and 2004. A total of 12 boreholes of potential relevance were identified. Further study showed that only four out of these holes penetrated into crystalline rock. Two of these were deemed to be less relevant because they were drilled in areas with much higher geothermal gradient than in the parts of the Fennoscandian shield that realistically could host a Swedish deep borehole repository. Of the two remaining boreholes, only one, a geoscientific hole drilled at Outokumpu in Finland, is associated with a reasonably complete geoscientific data set. It is worth mentioning that a large part of this hole is drilled through meta sedimentary rock (mica schist) rather than granitic rock. The information collected and reviewed has been gathered under the headings hydraulic conditions, geothermal conditions, hydrogeochemical conditions, bacteriological activity and rock mechanical properties. Only

Review of geoscientific data of relevance to disposal of spent nuclear fuel in deep boreholes in crystalline rock

Energy Technology Data Exchange (ETDEWEB)

Marsic, Nico; Grundfelt, Bertil [Kemakta Konsult AB, Stockholm (Sweden)

2013-09-15

In this report a compilation of recent geoscientific data of relevance to disposal of spent nuclear fuel in deep boreholes in Sweden is presented. The goal of the study has been limited to identifying and briefly describing such geoscientific information of relevance to disposal in deep boreholes that was not available at the time when previous compilations were made. Hence, the study is not to be regarded as a general up-date of new geoscientific information. Disposal of spent nuclear fuel in deep boreholes has been studied in Sweden since the second half of the 1980s. The currently studied concept has been proposed by Sandia National Laboratories in the USA. In this concept the spent fuel elements are encapsulated in cylindrical steel canisters that are joined together in strings of 40 canisters and lowered into five kilometres deep boreholes. Ten such strings are stacked between three and five kilometres depth separated from each other by concrete plugs. The study started with a review of boreholes that have been reported after the previous reviews that were published in 1998 and 2004. A total of 12 boreholes of potential relevance were identified. Further study showed that only four out of these holes penetrated into crystalline rock. Two of these were deemed to be less relevant because they were drilled in areas with much higher geothermal gradient than in the parts of the Fennoscandian shield that realistically could host a Swedish deep borehole repository. Of the two remaining boreholes, only one, a geoscientific hole drilled at Outokumpu in Finland, is associated with a reasonably complete geoscientific data set. It is worth mentioning that a large part of this hole is drilled through meta sedimentary rock (mica schist) rather than granitic rock. The information collected and reviewed has been gathered under the headings hydraulic conditions, geothermal conditions, hydrogeochemical conditions, bacteriological activity and rock mechanical properties. Only

As-Built Verification Plan Spent Nuclear Fuel Canister Storage Building MCO Handling Machine

International Nuclear Information System (INIS)

SWENSON, C.E.

2000-01-01

This as-built verification plan outlines the methodology and responsibilities that will be implemented during the as-built field verification activity for the Canister Storage Building (CSB) MCO HANDLING MACHINE (MHM). This as-built verification plan covers THE ELECTRICAL PORTION of the CONSTRUCTION PERFORMED BY POWER CITY UNDER CONTRACT TO MOWAT. The as-built verifications will be performed in accordance Administrative Procedure AP 6-012-00, Spent Nuclear Fuel Project As-Built Verification Plan Development Process, revision I. The results of the verification walkdown will be documented in a verification walkdown completion package, approved by the Design Authority (DA), and maintained in the CSB project files

The disposal of Canada's nuclear fuel waste: public involvement and social aspects

International Nuclear Information System (INIS)

Greber, M.A.; Frech, E.R.; Hillier, J.A.R.

1994-01-01

This report describes the activities undertaken to provide information to the public about the Canadian Nuclear Fuel Waste Management Program as well as the opportunities for public involvement in the direction and development of the disposal concept through government inquiries and commissions and specific initiatives undertaken by AECL. Public viewpoints and the major issues identified by the public to be of particular concern and importance in evaluating the acceptability of the concept are described. In addition, how the issues have been addressed during the development of the disposal concept or how they could be addressed during implementation of the disposal concept are presented. There is also discussion of public perspectives of risk, the ethical aspects of nuclear fuel waste disposal, and public involvement in siting a nuclear fuel waste disposal facility. The Canadian Nuclear Fuel Waste Management Program is funded jointly by AECL and Ontario Hydro under the auspices of the CANDU Owners Group. (author)

Thermal analysis of the horizontal disposal for HLW

International Nuclear Information System (INIS)

Zhao Honggang

2012-01-01

The temperature on the canister surface is set to be not more than 100 in the repository, a criterion which dictates the dimension of the repository. The factors that affect the highest temperature on the canister surface include the initial power of the canister, the material thermal properties of the engineered barrier system (EBS), the gaps around the canister in the EBS, the initial ground temperature and thermal properties of the host rock, the repository layout, etc. The article examines the material thermal properties of the host rock and the EBS, the thermal conductivity properties of the different gaps in the EBS, the temperature evolution around the single canister by using the analysis method and the numerical method for horizontal disposal concept. The findings are as follows: 1) The most important and the most sensitive parameter is the initial disposal power of the canister; 2) The two key factors that affect the highest temperature on the canister surface are the material parameter's uncertainty and nature variability of the host rock and the EBS, and the gaps around the canister in the EBS; 3) The temperature offsets between the canister and bentonite is not more than 10, and the bigger the inner gaps, the bigger temperature offsets between the canister and bentonite; When the gap between the bentonite and the host rock is filled with water, the gap's temperature offsets is small, but it will be 1∼3 higher when the gaps between the bentonite and the host rock is filled with air. (author)

Disposal of defense spent fuel and HLW from the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Ermold, L.F.; Loo, H.H.; Klingler, R.D.; Herzog, J.D.; Knecht, D.A.

1992-12-01

Acid high-level radioactive waste (HLW) resulting from fuel reprocessing at the Idaho Chemical Processing Plant (ICPP) for the US Department of Energy (DOE) has been solidified to a calcine since 1963 and stored in stainless steel bins enclosed by concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also in storage ate the ICPP. In April, 1992, DOE announced that spent fuel would no longer be reprocessed to recover enriched uranium and called for a shutdown of the reprocessing facilities at the ICPP. A new Spent Fuel and HLW Technology Development program was subsequently initiated to develop technologies for immobilizing ICPP spent fuels and HLW for disposal, in accordance with the Nuclear Waste Policy Act. The Program elements include Systems Analysis, Graphite Fuel Disposal, Other Spent Fuel Disposal, Sodium-Bearing Liquid Waste Processing, Calcine Immobilization, and Metal Recycle/Waste Minimization. This paper presents an overview of the ICPP radioactive wastes and current spent fuels, with an emphasis on the description of HLW and spent fuels requiring repository disposal

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

International Nuclear Information System (INIS)

Mitchell, D.E.

1986-01-01

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

State of the art of the welding method for sealing spent nuclear fuel canister made of copper. Part 2 - EBW

International Nuclear Information System (INIS)

Salonen, T.

2014-05-01

This report consist the results of the development of the electron beam welding (EBW) method for sealing spent nuclear fuel (SNF) disposal canister. This report has been used as background material for selection of the sealing method for the SNF canister. Report contains the state of the art knowledge of the EBW method and research and development (R and D) results done by Posiva. Relevant R and D results of EB-welds done by SKB are also reviewed in this report. Requirements set for the welding and weld are present. These requirements are based on the long term safety and also some part of requirements are set by other processes like non-destructive testing (NDT) and manufacturing processes of components. Initial state of the weld is described in this report. Initial state has significant effect on the long term safety issues like corrosion resistance and creep ductility. Also short and long term mechanical properties as well as corrosion properties are described. Microstructure and residual stresses of the weld is represented in this report. Report consists also imperfections of the weld and statistical analysis of the evaluation of the probability of the largest defect size on the weld. Results of corrosion and creep tests of EB-welds are reviewed in this report. EBW process and machine are described. Preliminary designing of the EBW-machine has been done including component handling equipments. Preliminary welding procedure specification (pWPS) has drawn up and qualification of the personnel is described briefly. In-line process and quality control system including seam tracking system is implemented in modern EBW machine. Also NDT methods for inspection of the weld are described in this report. Concerning the results from the research and development work it can be concluded that EB welding method is suitable method for sealing SNF canister. Weld material fulfils requirements set by the long term safety. The welding system is robust and reliable and it is based

Miniature Canister (MiniCan) Corrosion Experiment Progress Report 3 for 2008-2010

Energy Technology Data Exchange (ETDEWEB)

Smart, N.R.; Reddy, B.; Rance, A.P. (Serco (United Kingdom))

2011-08-15

To ensure the safe encapsulation of spent nuclear fuel rods for geological disposal, SKB of Sweden are considering using the Copper-Iron Canister, which consists of an outer copper canister and a cast iron insert. Over the years a programme of laboratory work has been carried out to investigate a range of corrosion issues associated with the canister, including the possibility of expansion of the outer copper canister as a result of the anaerobic corrosion of the cast iron insert. Previous experimental work using stacks of test specimens has not shown any evidence of corrosion-induced expansion. However, as a further step in developing an understanding of the likely performance of the canister in a repository environment, Serco has set up a series of experiments in SKB's Aespoe Hard Rock Laboratory (HRL) using inactive model canisters, in which leaks were deliberately introduced into the outer copper canister while surrounded by bentonite, with the aim of obtaining information about the internal corrosion evolution of the internal environment. The experiments use five small-scale model canisters (300 mm long x 150 mm diameter) that simulate the main features of the SKB canister design (hence the project name, 'MiniCan'). The main aim of the work is to examine how corrosion of the cast iron insert will evolve if a leak is present in the outer copper canister. This report describes the progress on the five experiments running at the Aespoe Hard Rock Laboratory and the data obtained from the start of the experiments in late 2006 up to Winter 2010. The full details of the design and installation of the experiments are given in a previous report and this report concentrates on summarising and interpreting the data obtained to date. This report follows two earlier progress reports presenting results up to December 2009. The current document (progress report 3) describes work up to December 2010. The current report presents the results of the water analyses

Disposal of radioactive waste in Swedish crystalline rocks

International Nuclear Information System (INIS)

Greis Dahlberg, Christina; Wikberg, Peter

2015-01-01

SKB, Swedish Nuclear Fuel and Waste Management Company is tasked with managing Swedish nuclear and radioactive waste. Crystalline rock is the obvious alternative for deep geological disposal in Sweden. SKB is, since 1988, operating a near surface repository for short-lived low and intermediate-level waste, SFR. The waste in SFR comprises operational and decommissioning waste from nuclear plants, industrial waste, research-related waste and medical waste. Spent nuclear fuel is currently stored in an interim facility while waiting for a license to construct a deep geological repository. The Swedish long-lived low and intermediate-level waste consists mainly of BWR control rods, reactor internals and legacy waste from early research in the Swedish nuclear programs. The current plan is to dispose of this waste in a separate deep geological repository, SFL, sometimes after 2045. Understanding of the rock properties is the basis for the design of the repository concepts. Swedish crystalline rock is mechanical stable and suitable for underground constructions. The Spent Fuel Repository is planned at approximately 500 meters depth in the rock at the Forsmark site. The host rock will keep the spent fuel isolated from human and near-surface environment. The rock will also provide the stable chemical and hydraulic conditions that make it possible to select suitable technical barriers to support the containment provided by the rock. A very long lasting canister is necessary to avoid release and transport of radionuclides through water conducting fractures in the rock. A canister designed for the Swedish rock, consists of a tight, 5 cm thick corrosion barrier of copper and a load-bearing insert of cast iron. To restrict the water flow around the canister and by that prevent fast corrosion, a bentonite buffer will surround the canister. Secondary, the bentonite buffer will retard a potential release by its strong sorption of radionuclides. The SFR repository is situated in

Disposal of radioactive waste in Swedish crystalline rocks

Energy Technology Data Exchange (ETDEWEB)

Greis Dahlberg, Christina; Wikberg, Peter [Svensk Kaernbraenslehantering AB, Stockholm (Sweden)

2015-07-01

SKB, Swedish Nuclear Fuel and Waste Management Company is tasked with managing Swedish nuclear and radioactive waste. Crystalline rock is the obvious alternative for deep geological disposal in Sweden. SKB is, since 1988, operating a near surface repository for short-lived low and intermediate-level waste, SFR. The waste in SFR comprises operational and decommissioning waste from nuclear plants, industrial waste, research-related waste and medical waste. Spent nuclear fuel is currently stored in an interim facility while waiting for a license to construct a deep geological repository. The Swedish long-lived low and intermediate-level waste consists mainly of BWR control rods, reactor internals and legacy waste from early research in the Swedish nuclear programs. The current plan is to dispose of this waste in a separate deep geological repository, SFL, sometimes after 2045. Understanding of the rock properties is the basis for the design of the repository concepts. Swedish crystalline rock is mechanical stable and suitable for underground constructions. The Spent Fuel Repository is planned at approximately 500 meters depth in the rock at the Forsmark site. The host rock will keep the spent fuel isolated from human and near-surface environment. The rock will also provide the stable chemical and hydraulic conditions that make it possible to select suitable technical barriers to support the containment provided by the rock. A very long lasting canister is necessary to avoid release and transport of radionuclides through water conducting fractures in the rock. A canister designed for the Swedish rock, consists of a tight, 5 cm thick corrosion barrier of copper and a load-bearing insert of cast iron. To restrict the water flow around the canister and by that prevent fast corrosion, a bentonite buffer will surround the canister. Secondary, the bentonite buffer will retard a potential release by its strong sorption of radionuclides. The SFR repository is situated in

Decision nearing on final disposal of spent fuel in Finland

International Nuclear Information System (INIS)

Vira, J.

2000-01-01

The programme for final disposal of spent fuel from Finnish nuclear power plants is entering into important phase: in the year 2000 the Finnish Government is expected to decide whether the proposal made by Posiva Oy on the spent fuel disposal is in line with the overall good of society. Associated with the decision is also Posiva's proposal on siting the disposal facility at Olkiluoto in Eurajoki municipality on the western coast of Finland. An important document underlying Posiva's application for this principle decision is the report of the environmental impact assessment, which was completed in 1999. Safety considerations play an important role in the application. New assessments have, therefore, been made on both the operational and long-term safety as well as on safety of spent fuel transportation. (author)

Magnox fuel dry storage and direct disposal assessment of CEGB/SSEB reports

International Nuclear Information System (INIS)

1987-12-01

This report assesses the Boards' presented work in response to Recommendations 17 and 18 of the Environment Committee's First Report (Jan 86). The Boards have made an extensive study of the dry store design and also considered direct disposal. Their basic conclusion that the financial advantage is with continued reprocessing is accepted with the comment that their storage and disposal costs may be on the high side. The Boards statements on drying wet-stored fuel and on improvement of the fuel's chemical stability are accepted. The Boards coverage of fuel after disposal is considered to be too brief; the assessment expresses a more pessimistic view than the Boards' of the acceptability of direct disposal. (author)

Multi-purpose canister storage unit and transfer cask thermal analysis

International Nuclear Information System (INIS)

Montgomery, R.A.; Niemer, K.A.; Lindner, C.N.

1997-01-01

Spent Nuclear Fuel (SNF) generated at commercial nuclear power plants throughout the US is a concern because of continued delays in obtaining a safe, permanent disposal facility. Most utilities maintain their SNF in wet storage pools; however, after decades of use, many pools are filled to capacity. Unfortunately, DOE's proposed final repository at Yucca Mountain is at least 10 years from completion, and commercial power utilities have few options for SNF storage in the interim. The Multi-Purpose Canister (MPC) system, sponsored by DOE's Office of Civilian Radioactive Waste Management, is a viable solution to the interim storage problem. The system is designed for interim dry storage, transport, and ultimate disposal of commercial SNF. The MPC system consists of four separate components: an MPC, Transfer Cask, Storage Unit, and Transport Cask. The SNF assemblies are loaded and sealed inside the helium-filled steel MPC. Once sealed, the MPC is not reopened, eliminating the need to re-handle the individual spent fuel assemblies. The MPC is transferred, using the MPC Transfer Cask, into a cylindrical, reinforced-concrete Storage Unit for on-site dry storage. The MPC may be removed from the Storage Unit at any time and transferred into the MPC Transport Cask for transport to the final repository. This paper discusses the analytical approach used to evaluate the heat transfer characteristics of an MPC containing SNF assemblies in the MPC Transfer Cask and Storage Unit