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Sample records for spent resin transport

  1. Bituminous solidification, disposal, transport and burial of spent ion-exchange resins. Part of a coordinated programme on treatment of spent ion exchange resins

    International Nuclear Information System (INIS)

    Mozes, G.; Kristof, M.

    1983-07-01

    The project dealing with the incorporation of spent ion-exchange resins into bitumen was performed within the Agency coordinated research programme on treatment of spent ion-exchange resins. Physical and chemical properties of commercial ion-exchange resins, bitumens and bituminized resins were studied. It was shown that bitumen with low oil content and with a softening point of 60-70 deg. C are applicable for the incorporation of resins. The final waste form is allowed to contain maximum 50% resin. The comprehensive study of the biological resistance of B-30 bitumen was performed. That showed that any bacteriological attack can be regarded as generally insignificant. A continuously operating technology was realized on a semi-plant scale. The best operating conditions of this technology were determined. On the basis of the experience gained from the experiments a design of the bituminization plant of 50m 3 dry resin/year treatment capacity was proposed

  2. Immobilization of spent resin with epoxy resin

    International Nuclear Information System (INIS)

    Gultom, O.; Suryanto; Sayogo; Ramdan

    1997-01-01

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

  3. The solidification of spent resin

    International Nuclear Information System (INIS)

    Shiao, S. J.; Tsai, C. M.; Shyu, Y. H.

    1991-01-01

    A quasi-steady apparatus was applied to measure the thermal conductivity of solids ranging in size for 0.3 to 200 L, and temperature distributions in the solids were recorded during the curing, and theoretical equation for conduction in a cylindrical form with uniform energy generation was established to define the thermal state of reaction. The heat of reaction calculated from the theoretical equation with experimental values for the maximum temperature and thermal conductivity agrees very well with the data reported. The relationships among heat of reaction and amount of curing agent, retardant, loading of spent resin, and water were established

  4. Electrodialytic decontamination of spent ion exchange resins

    International Nuclear Information System (INIS)

    Nott, B.R.

    1982-01-01

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

  5. Processing method of radioactive spent ion exchange resin

    International Nuclear Information System (INIS)

    Shimizu, Takayuki; Saito, Tomohisa; Kurosaka, Norio.

    1997-01-01

    Resin beads and resin powders are collected from respective reservoir tanks to a supply tank, and the resin beads and the resin powders are continuously transferred from the supply tank to a dehydrator depending on the processing performance thereof to conduct dehydration. The obtained dehydrated resin beads and the powders are formed into a solidification product by cement solidification, or by incineration followed by cement solidification of incineration ashes, or are temporary stored in a storage vessel, and then subjected to one of the solidification treatments. It is possible to eliminate a worry of forming lumpy spent ion exchange resins in a storage tank to cause troubles upon withdrawal from the tank or during transportation. (N.H.)

  6. Spent fuel transportation problems

    International Nuclear Information System (INIS)

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

    1977-01-01

    In this paper, problems of transportation of nuclear spent fuel to reprocessing plants are discussed. The solutions proposed are directed toward the achievement of the transportation as economic and safe as possible. The increase of the nuclear power plants number in the USSR and the great distances between these plants and the reprocessing plants involve an intensification of the spent fuel transportation. Higher burnup and holdup time reduction cause the necessity of more bulky casks. In this connection, the economic problems become still more important. One of the ways of the problem solution is the development of rational and cheap cask designs. Also, the enforcement in the world of the environmental and personnel health protection requires to increase the transportation reliability and safety. The paper summarizes safe transportation rules with clarifying the following questions: the increase of the transport unit quantity of the spent fuel; rational shipment organization that minimizes vehicle turnover cycle duration; development of the reliable calculation methods to determine strength, thermal conditions and nuclear safety of transport packaging as applied to the vehicles of high capacity; maximum unification of vehicles, calculation methods and documents; and cask testing on models and in pilot scale on specific test rigs to assure that they meet the international safe fuel shipment rules. Besides, some considerations on the choice and use of structural materials for casks are given, and problems of manufacturing such casks from uranium and lead are considered, as well as problems of the development of fireproof shells, control instrumentation, vehicles decontamination, etc. All the problems are considered from the point of view of normal and accidental shipment conditions. Conclusions are presented [ru

  7. Processing ix spent resin waste for C-14 isotope recovery

    International Nuclear Information System (INIS)

    Chang, F. H.; Woodall, K. B.; Sood, S. K.; Vogt, H. K.; Krochmainek, L. S.

    1991-01-01

    A process developed at Ontario Hydro for recovering carbon-14 (C-14) from spent ion exchange resin wastes is described. Carbon-14 is an undesirable by-product of CANDU 1 nuclear reactor operation. It has an extremely long (5730 years) half-life and can cause dosage to inhabitants by contact, inhalation, or through the food cycle via photosynthesis. Release of carbon-14 to the environment must be minimized. Presently, all the C-14 produced in the Moderator and Primary Heat Transport (PHT) systems of the reactor is effectively removed by the respective ion exchange columns, and the spent ion exchange resins are stored in suitably engineered concrete structures. Because of the large volumes of spent resin waste generated each year this method of disposal by long term storage tends to be uneconomical; and may also be unsatisfactory considering the long half-life of the C-14. However, purified C-14 is a valuable commercial product for medical, pharmaceutical, agricultural, and organic chemistry research. Currently, commercial C-14 is made artificially in research reactors by irradiating aluminum nitride targets for 4.5 years. If the C-14 containing resin waste can be used to reduce this unnecessary production of C-14, the total global build-up of this radioactive chemical can be reduced. There is much incentive in removing the C-14 from the resin waste to reduce the volume of C-14 waste, and also in purifying the recovered C-14 to supply the commercial market. The process developed by Ontario Hydro consists of three main steps: C-14 removal from spent resins, enrichment of recovered C-14, and preparation of final product. Components of the process have been successfully tested at Ontario Hydro's Research Division, but the integration of the process is yet to be demonstrated. A pilot scale plant capable of processing 4 m 3 of spent resins annually is being planned for demonstrating the technology. The measured C-14 activity levels on the spent resins ranged from 47

  8. Spent fuel transporting vessel

    International Nuclear Information System (INIS)

    Kumagaya, Takeshi.

    1995-01-01

    A large number of annular cooling fins are disposed each at an equal distance on the outer circumferential surface of a vessel main body. An electric power generation module is disposed on the surface of the cooling fins. The electric power generation module comprises a plurality of thermoelectric power generation elements. In each of the thermoelectric generation elements, the inner side thereof in contact with the surface of the cooling fin is at a high temperature while the outer side thereof is at a low temperature nearly equal with an atmospheric temperature. A predetermined amount of electric power is generated by seebeck effect due to the temperature difference. The electric power is always stored in a battery. Accordingly, even if a power generator of a ship should fail and power supply is stopped during transportation of the vessels for spent nuclear fuels, an appropriate amount of electric power can be supplied to a cooling device of the ship. (I.N.)

  9. Transportation of spent nuclear fuels

    International Nuclear Information System (INIS)

    Meguro, Toshiichi

    1976-01-01

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

  10. Intermodal transportation of spent fuel

    International Nuclear Information System (INIS)

    Elder, H.K.

    1983-09-01

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

  11. Worldwide spent fuel transportation logistics

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  12. Pilot Plant for treating of spent exchange resins

    International Nuclear Information System (INIS)

    Iglesias, Alberto M.; Raffo Calderon, Maria del C.; Varani, Jose L.

    2004-01-01

    Spent exchange resins that have been accumulating during the last operational 30 years in Atucha I nuclear power plant (NPP) are a 'problematic' waste. These spent resins conform an intermediate level waste due to the total content of alpha, beta and gamma emitters (some samples of spent resins were analyzed in 2003). For this reason its treatment is more expensive since it is necessary to add more safety barriers for its final disposition and also for the radioprotection actions that are involved. Using sulfuric acid solutions it is possible to elute from the spent resins the ions that are retained. In the same operation are eluted Cobalt, Cesium and alpha emitters since that all these elements react as cations in aqueous solution. Decontamination by electrochemical methods was analyzed as an interesting method to apply after elution operation to these spent resins since that with the decontamination process it is possible to obtain a solid without activity and concentrate the activity in cells that are small in volume and its manipulation doesn't present any extra complication. Experiments made with active samples taken from the deposit were successful. Because of these results it was built a small plant to treat a batch of 100 dm 3 of wet spent exchange resins. Some problems with the material that was in the deposit together with spent resins caused that we had to plan a more complex strategy to obtain a complete decontamination of the spent resins (in this stage we used the cobalt retention cell that was described in other paper to retain Cobalt and alpha emitters and a sample of zeolites from Argentina ores to retain Cesium). Due to alpha emitters act electrochemically like cations it was possible to retain altogether with ionic Cobalt on the copper amalgam electrode. Working in the non-active lab with alcoholic solutions it was possible to retain ionic Cesium on a copper electrode (copper is covered by mercury fine film which forms a solid amalgam) with a

  13. Immobilisation of spent ion exchange resin in polyester-styrene co-polymer

    International Nuclear Information System (INIS)

    Kumar, Amar

    1999-01-01

    Spent ion exchange resin is generated from primary heat transport system, moderator system and spent fuel storage bay of PHWR which contain large amounts of activated corrosion products and fission products. It was immobilised in crack free monolith of polyester styrene co-polymer. 1.0% of benzoyl peroxide and 0.07% of dimethyl aniline were used as initiator and activator respectively. Gelling time of the polymerisation was 30 minutes. Product has good chemical, mechanical and radiation stability. (author)

  14. Spent nuclear fuel transport problems

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  15. Carbonization-cementation process for treatment of spent IX resins

    International Nuclear Information System (INIS)

    Sathi Sasidharan, N.; Deshingkar, D.S.; Wattal, P.K.

    2004-08-01

    The spent IX resins containing radioactive fission and activation products of reactor structural materials are highly active solid wastes generated during operations of nuclear reactors. Feasibility tests were conducted for carbonization of IX resins to achieve weight and volume reduction and destruction of functional groups so as to make them compatible for immobilization in cement matrix. Carbonization of non-radioactive resins was studied at 250 to 350 degC. Carbonization residues were 20 to 32 wt% depending upon the type of resin and temperature of carbonization. The release of 137 Cs activity to off-gases was 0.004% at 300 degC and 0.05 % at 350 degC. Based on these tests, a 50 liter/batch capacity inactive resin carbonization pilot plant was set up. Carbonization residues could be immobilized into cement matrix with 60 wt % loading using vermiculite and precipitated silica as admixtures. The cumulative fraction of 137 Cs leached from the selected cement matrix was 0.0066 in 200 days. Based on pilot plant studies and cementation tests, the swollen spent resins waste volume could be minimized by 2.7 times. (author)

  16. Probability of spent fuel transportation accidents

    International Nuclear Information System (INIS)

    McClure, J.D.

    1981-07-01

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

  17. Immobilisation Of Spent Ion Exchange Resins Using Portland Cement Blending With Organic Material

    International Nuclear Information System (INIS)

    Zalina Laili; Mohd Abdul Wahab; Nur Azna Mahmud

    2014-01-01

    Immobilisation of spent ion exchange resins (spent resins) using Portland cement blending with organic material for example bio char was investigated. The performance of cement-bio char matrix for immobilisation of spent ion exchange resins was evaluated based on their compression strength and leachability under different experimental conditions. The results showed that the amount of bio char and spent resins loading effect the compressive strength of the waste form. Several factors affecting the leaching behaviour of immobilised spent resins in cement-bio char matrix. (author)

  18. Spent fuels transportation coming from Australia

    International Nuclear Information System (INIS)

    2002-01-01

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

  19. Ontario Hydro Research Division's program for treatment of spent ion-exchange resins

    International Nuclear Information System (INIS)

    Nott, B.R.; Dodd, D.J.R.

    1981-09-01

    A brief review of the evolution of work programmes for chemical treatment of spent ion-exchange resins in Ontario Hydro's Research Division is presented. Attention has been focussed on pre-treatment processes for the treatment of the spent resins prior to encapsulation of the products in solid matrices. Spent Resin Regeneration and Acid Stripping processes were considered in some detail. Particular attention was paid to carbon-14 on spent resins, its determination in and removal from the spent resins (with the acid stripping technique). The use of separate cation and anion resin beds instead of mixed bed resins was examined with a view to reducing the volume of resin usage and consequently the volume of waste radioactive ion-exchange resin generated. (author)

  20. Degradation of ion spent resin using the Fenton's reagent

    International Nuclear Information System (INIS)

    Araujo, Leandro Goulart de

    2013-01-01

    The most common method for spent radioactive ion exchange resin treatment is its immobilization in cement, which reduces the radionuclides release into the environment. Although this method is efficient, it increases considerably the final volume of the waste due to the low incorporation capacity. The objective of this work was to develop a degradation method of spent resins arising from the nuclear research reactor located at the Nuclear and Energy Research Institute (IPEN-CNEN/SP), using an Advanced Oxidation Process (AOP) with Fenton's reagents. This method would allow a higher incorporation in cement. Three different resins were evaluated: cationic, anionic and a mixture of both resins. The reactions were conducted varying the catalyst concentration (25, 50, 100 and 150 mM), the volume of hydrogen peroxide (320 to 460 mL), and three different temperatures, 50, 60 and 70 deg C. Degradation of about 98% was achieved using a 50 mM catalyst solution and 330 mL of hydrogen peroxide solution. The most efficient temperature was 60 deg C. (author)

  1. Thermal model of spent fuel transport cask

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  2. Cation immobilization in pyrolyzed simulated spent ion exchange resins

    Energy Technology Data Exchange (ETDEWEB)

    Luca, Vittorio, E-mail: vluca@cnea.gov.ar [Programa Nacional de Gestion de Residuos Radiactivos, Centro Atomico Constituyentes, Comision Nacional de Energia Atomica, Av. General, Paz 1499, 1650 San Martin, Provincia de Buenos Aires (Argentina); Bianchi, Hugo L. [Gerencia de Quimica, Centro Atomico Constituyentes, Comision Nacional de Energia Atomica, Av. General, Paz 1499, 1650 San Martin, Provincia de Buenos Aires (Argentina); ECyT, Universidad Nacional de General San Martin, Campus Miguelete, Ed. Tornavias, Martin de Irigoyen 3100, 1650 San Martin (Argentina); Conicet, Av. Rivadavia 1917, 1033 Buenos Aires (Argentina); Manzini, Alberto C. [Programa Nacional de Gestion de Residuos Radiactivos, Comision Nacional de Energia Atomica, Av. Del Libertador 8250, CP 1429, Ciudad Autonoma de Buenos Aires (Argentina)

    2012-05-15

    Significant quantities of spent ion exchange resins that are contaminated by an assortment of radioactive elements are produced by the nuclear industry each year. The baseline technology for the conditioning of these spent resins is encapsulation in ordinary Portland cement which has various shortcomings none the least of which is the relatively low loading of resin in the cement and the poor immobilization of highly mobile elements such as cesium. The present study was conducted with cationic resin samples (Lewatit S100) loaded with Cs{sup +}, Sr{sup 2+}, Co{sup 2+}, Ni{sup 2+} in roughly equimolar proportions at levels at or below 30% of the total cation exchange capacity. Low temperature thermal treatment of the resins was conducted in inert (Ar), or reducing (CH{sub 4}) gas atmospheres, or supercritical ethanol to convert the hydrated polymeric resin beads into carbonaceous materials that contained no water. This pyrolytic treatment resulted in at least a 50% volume reduction to give mechanically robust spherical materials. Scanning electron microscope investigations of cross-sections of the beads combined with energy dispersive analysis showed that initially all elements were uniformly distributed through the resin matrix but that at higher temperatures the distribution of Cs became inhomogeneous. Although Cs was found in the entire cross-section, a significant proportion of the Cs occurred within internal rings while a proportion migrated toward the outer surfaces to form a crustal deposit. Leaching experiments conducted in water at 25 Degree-Sign C showed that the divalent contaminant elements were very difficult to leach from the beads heated in inert atmospheres in the range 200-600 Degree-Sign C. Cumulative fractional loses of the order of 0.001 were observed for these divalent elements for temperatures below 500 Degree-Sign C. Regardless of the processing temperature, the cumulative fractional loss of Cs in water at 25 Degree-Sign C reached a plateau or

  3. Cation immobilization in pyrolyzed simulated spent ion exchange resins

    International Nuclear Information System (INIS)

    Luca, Vittorio; Bianchi, Hugo L.; Manzini, Alberto C.

    2012-01-01

    Significant quantities of spent ion exchange resins that are contaminated by an assortment of radioactive elements are produced by the nuclear industry each year. The baseline technology for the conditioning of these spent resins is encapsulation in ordinary Portland cement which has various shortcomings none the least of which is the relatively low loading of resin in the cement and the poor immobilization of highly mobile elements such as cesium. The present study was conducted with cationic resin samples (Lewatit S100) loaded with Cs + , Sr 2+ , Co 2+ , Ni 2+ in roughly equimolar proportions at levels at or below 30% of the total cation exchange capacity. Low temperature thermal treatment of the resins was conducted in inert (Ar), or reducing (CH 4 ) gas atmospheres, or supercritical ethanol to convert the hydrated polymeric resin beads into carbonaceous materials that contained no water. This pyrolytic treatment resulted in at least a 50% volume reduction to give mechanically robust spherical materials. Scanning electron microscope investigations of cross-sections of the beads combined with energy dispersive analysis showed that initially all elements were uniformly distributed through the resin matrix but that at higher temperatures the distribution of Cs became inhomogeneous. Although Cs was found in the entire cross-section, a significant proportion of the Cs occurred within internal rings while a proportion migrated toward the outer surfaces to form a crustal deposit. Leaching experiments conducted in water at 25 °C showed that the divalent contaminant elements were very difficult to leach from the beads heated in inert atmospheres in the range 200–600 °C. Cumulative fractional loses of the order of 0.001 were observed for these divalent elements for temperatures below 500 °C. Regardless of the processing temperature, the cumulative fractional loss of Cs in water at 25 °C reached a plateau or steady-state within the first 24 h increasing only

  4. Treatment of spent ion-exchange resins for storage and disposal

    International Nuclear Information System (INIS)

    1985-01-01

    This report describes the experience gained by different countries on storage of spent ion exchange resins, immobilization of them into various matrices and the development of new methods in decomposition and solidification of spent resins. The report contains all the results of the Coordinated Research Programme together with additional data available from countries not participating in this programme. A review of practical industrial experience in treating spent ion exchange resins is given in the annex

  5. Transporting spent nuclear fuel: an overview

    International Nuclear Information System (INIS)

    1986-03-01

    Although high-level radioactive waste from both commercial and defense activities will be shipped to the repository, this booklet focuses on various aspects of transporting commercial spent fuel, which accounts for the majority of the material to be shipped. The booklet is intended to give the reader a basic understanding of the following: the reasons for transportation of spent nuclear fuel, the methods by which it is shipped, the safety and security precautions taken for its transportation, emergency response procedures in the event of an accident, and the DOE program to develop a system uniquely appropriate to NWPA transportation requirements

  6. Study of mechanical and physicochemical properties of cementated spent ion-exchange-resins

    International Nuclear Information System (INIS)

    Patek, P.

    1981-09-01

    As first part of a study on the possibilities, to immobilize spent ion exchange resins, for final disposal, the dependence of compressive strength from the composition of cement - resin mixtures was detected. Powdered resins, bead resins and ashes from the incinerator plant and several cement brands were examinated. As result an area was defined in the three-phase diagram of cement, resins and water, in which the following leach tests will be performed. (author)

  7. Leaching of 60 Co and 137 Cs from spent ion exchange resins in ...

    Indian Academy of Sciences (India)

    Cement; radioactive waste; composite; waste management. Abstract. The leaching rate of 60Co and 137Cs from the spent cation exchange resins in cement–bentonite matrix has been studied. The solidification matrix was a standard Portland cement mixed with 290–350 (kg/m3) spent cation exchange resins, with or ...

  8. Considerations for the transportation of spent fuel

    International Nuclear Information System (INIS)

    Jefferson, R.M.

    1984-01-01

    In our society today the transportation of radioactive materials, and most particularly spent reactor fuel, is surrounded by considerable emotion and a wealth of information, good and bad. The transportation of these materials is viewed as unique and distinct from the transportation of other hazardous materials and as a particularly vulnerable component of the nuclear power activities of this nation. Added to this is the concept, widely held, that almost everyone is an expert on the transportation of radioactive materials. One significant contribution to this level of emotion is the notion that all roads (rail and highway), on which these goods will be transported, somehow traverse everyone's backyard. The issue of the transportation of spent fuel has thus become a political battleground. Perhaps this should not be surprising since it has all of the right characteristics for such politicization in that it is pervasive, emotional, and visible. In order that those involved in the discussion of this activity might be able to reach some rational conclusions, this paper offers some background information which might be useful to a broad range of individuals in developing their own perspectives. The intent is to address the safety of transporting spent fuel from a technical standpoint without the emotional content which is frequently a part of this argument

  9. Transportation 2000. Spent fuel transportation trends in the new millenium

    International Nuclear Information System (INIS)

    Blee, David; Viebrock, James; Patterson, John

    1999-01-01

    The paper will provide a comparison of foreign research reactor spent fuel transportation today verses the assumptions used by the Department of Energy in the Environmental Impact Statement. In addition, it will suggest changes that are likely to occur in transportation logistics through the remainder of the U.S. spent fuel returns program. Cask availability, certification status, shipment strategy, cost issues, and public acceptance are among the topical areas that will be examined. Transportation requirements will be assessed in light of current participation in the returns program and the tendency for shipment plans to shift toward spent fuel return toward the end of the 13 year period of eligibility. (author)

  10. Transport and storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Lung, M.; Lenail, B.

    1987-01-01

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

  11. European experience with spent fuel transport

    International Nuclear Information System (INIS)

    Hunter, I.A.

    1995-01-01

    Nuclear Transport Ltd has transported 5000 tonnes of spent fuel from 35 reactors in 8 European countries since 1972. Transport management is governed by the Quality Plan for: transport administration, packaging and shipment procedures at the shipping plant, operations at the power plant, and packaging and shipment organization at the power plant. Selection of a suitable carrier device is made with regard to the shipping plant requirements, physical limitations of the reactor, fuel characteristics, and transport route constraints. The transport plan is set up taking into account exploitation of the casks, reactor shut-down requirements, fuel acceptance plans at the reprocessing plant, and cask maintenance periods. A transport cycle involving spent fuel shipment to La Hague or to Sellafield takes typically two or four weeks, respectively. Most transports through Europe are by rail. A special-design railway ferry boat serves transports to the United Kingdom. Both wet or dry casks are employed. Modern casks are designed for high burnups and for oxide fuels. (J.B.)

  12. Nuclear fuel transport and particularly spent fuel transport

    International Nuclear Information System (INIS)

    Lenail, B.

    1986-01-01

    Nuclear material transport is an essential activity for COGEMA linking the different steps of the fuel cycle transport systems have to be safe and reliable. Spent fuel transport is more particularly examined in this paper because the development of reprocessing plant. Industrial, techmical and economical aspects are reviewed [fr

  13. The characteristic assessment of spent ion exchange resin from PUSPATI TRIGA REACTOR (RTP) for immobilization process

    Energy Technology Data Exchange (ETDEWEB)

    Wahida, Nurul [School of Applied Physics, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor, Malaysia and Malaysian Nuclear Agency, Bangi 43000 Kajang, Selangor (Malaysia); Yasir, Muhamad Samudi; Majid, Amran Ab; Irwan, M. N. [School of Applied Physics, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor (Malaysia); Wahab, Mohd Abd; Marzukee, Nik; Paulus, Wilfred; Phillip, Esther; Thanaletchumy [Malaysian Nuclear Agency, Bangi 43000 Kajang, Selangor (Malaysia)

    2014-09-03

    In this paper, spent ion exchange resin generated from PUSPATI TRIGA reactor (RTP) in Malaysian Nuclear Agency were characterized based on the water content, radionuclide content and radionuclide leachability. The result revealed that the water content in the spent resin is 48%. Gamma spectrometry analysis indicated the presence of {sup 134}Cs, {sup 137}Cs, {sup 152}Eu, {sup 54}Mn, {sup 58}Co, {sup 60}Co and {sup 65}Zn. The leachability test shows a small concentrations (<1 Bq/l) of {sup 152}Eu and {sup 134}Cs were leached out from the spent resin while {sup 60}Co activity concentrations slightly exceeded the limit generally used for industrial wastewater i.e. 1 Bq/l. Characterization of spent ion exchange resin sampled from RTP show that this characterization is important as a basis to immobilize this radioactive waste using geopolymer technology.

  14. Catalytic oxidative pyrolysis of spent organic ion exchange resins from nuclear power plants

    International Nuclear Information System (INIS)

    Sathi Sasidharan, N.; Deshingkar, D.S.; Wattal, P.K.; Shirsat, A.N.; Bharadwaj, S.R.

    2005-08-01

    The spent IX resins from nuclear power reactors are highly active solid wastes generated during operations of nuclear reactors. Catalytic oxidative pyrolysis of these resins can lead to high volume reduction of these wastes. Low temperature pyrolysis of transition metal ion loaded IX resins in presence of nitrogen was carried out in order to optimize catalyst composition to achieve maximum weight reduction. Thermo gravimetric analysis of the pyrolysis residues was carried out in presence of air in order to compare the oxidative characteristics of transition metal oxide catalysts. Copper along with iron, chromium and nickel present in the spent IX resins gave the most efficient catalyst combination for catalytic and oxidative pyrolysis of the residues. During low temperature catalytic pyrolysis, 137 Cesium volatility was estimated to be around 0.01% from cationic resins and around 0.1% from anionic resins. During oxidative pyrolysis at 700 degC, nearly 10 to 40% of 137 Cesium was found to be released to off gases depending upon type of resin and catalyst loaded on to it. The oxidation of pyrolytic residues at 700 degC gave weight reduction of 15% for cationic resins and 93% for anionic resins. Catalytic oxidative pyrolysis is attractive for reducing weight and volume of spent cationic resins from PHWRs and VVERs. (author)

  15. Research and development for decontamination system of spent resin in Hanbit Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Gi Hong [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2015-12-15

    When reactor coolant leaks occur due to cracks of a steam generator tube, radioactive materials contained in the primary cooling water in nuclear power plant are forced out toward the secondary systems. At this time the secondary water purification resin in the ion exchange resin tower of the steam generator blowdown system is contaminated by the radioactivity of the leaked radioactive materials, so we pack this in special containers and store temporarily because we could not dispose it by ourselves. If steam generator tube leakage occurs, it produces contaminated spent resins annually about 5,000-7,000 liters. This may increase the amount of nuclear waste productions, a disposal working cost and a unit price of generating electricity in the plant. For this reasons, it is required to develop a decontamination process technique for reducing the radioactive level of these resins enough to handle by the self-disposal method. In this research, First, Investigated the structure and properties of the ion exchange resin used in a steam generator blowdown system. Second, Checked for a occurrence status of contaminated spent resin and a disposal technology. Third, identified the chemical characteristics of the waste radionuclides of the spent resin, and examined ionic bonding and separation mechanism of radioactive nuclear species and a spent resin. Finally, we carried out the decontamination experiment using chemicals, ultrasound, microbubbles, supercritical carbon dioxide to process these spent resin. In the case of the spent resin decontamination method using chemicals, the higher the concentration of the drug decontamination efficiency was higher. In the ultrasound method, foreign matter of the spent resin was removed and was found that the level of radioactivity is below of the MDA. In the microbubbles method, we found that the concentration of the radioactivity decreased after the experiment, so it can be used to the decontamination process of the spent resin. In

  16. Remarks on the transportation of spent fuel elements

    International Nuclear Information System (INIS)

    Krull, W.

    1992-01-01

    Information and data are provided on several aspects of the transportation of spent fuel elements. These aspects include contract, transportation, reprocessing batch size, and economical considerations. (author)

  17. Evaluation of the Compressive Strength of Cement-Spent Resins Matrix Mixed with Bio char

    International Nuclear Information System (INIS)

    Zalina Laili; Muhamad Samudi Yasir; Zalina Laili; Mohd Abdul Wahab; Nur Azna Mahmud; Nurfazlina Zainal Abidin

    2015-01-01

    The evaluation of compressive strength of cement-spent resins matrix mixed with bio char was investigated. In this study, bio char with different percentage (5 %, 8 %, 11 % 14 % and 18 %) was used as alternative admixture material for cement solidification of spent resins. Some properties of the physical and chemical of spent resins and bio char were also investigated. The performance of cemented spent resins with the addition of bio char was evaluated based on their compressive strength and the water resistance test. The compressive strength was evaluated at three different curing periods of 7, 14 and 28 days, while 4 weeks of immersion in distilled water was chosen for water resistance test. The result indicated that the compressive strength at 7, 14 and 28 days of curing periods were above the minimum criterion for example > 3.45 MPa of acceptable level for cemented waste form. Statistical analysis showed that there was no significant relationship between the compressive strength of the specimen and the percentage of bio char content. Result from the water resistance test showed that only one specimen that contained of 5 % of bio char failed the water resistance test due to the high of spent resins/ bio char ratio. The compressive strength of cement solidified spent resins was found increased after the water resistance test indicating further hydration occurred after immersed in water. The results of this study also suggest that the specimen with 8 %, 11 %, 14 % and 18 % of bio char content were resistance in water and suitable for the leaching study of radionuclides from cement-bio char-spent resins matrix. (author)

  18. Nevada commercial spent nuclear fuel transportation experience

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-09-01

    The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed.

  19. Nevada commercial spent nuclear fuel transportation experience

    International Nuclear Information System (INIS)

    1991-09-01

    The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed

  20. Solidification of spent ion exchange resin using new cementitious material, (1)

    International Nuclear Information System (INIS)

    Matsuda, Masami; Nishi, Takashi; Chino, Koichi; Kikuchi, Makoto

    1992-01-01

    Cement solidification is a conventional technique to solidify spent ion exchange resin. Swelling property of the ion exchange resin was investigated theoretically and experimentally, as the first step to improve resin content in a cement package. Swelling pressure of the resin was first formulated based on Gregor's swelling model. It was shown that the pressure depended on such factors as ion exchange capacity, type of functional group, crosslinkage and water content. Subsequently, swelling pressures for over forty properties of cation exchange resin (DIAION SK-series) with different exchange capacities, water contents and others were experimentally evaluated using an originally developed method. The pressure changed over a range of 0 ∼ 60 MPa, depending on properties of the ion exchange resin. These results were in close agreement with values calculated by Gregor's model, indicating that the swelling pressure could be evaluated without resorting to experiments. (author)

  1. Decontamination of spent ion-exchangers contaminated with cesium radionuclides using resorcinol-formaldehyde resins.

    Science.gov (United States)

    Palamarchuk, Marina; Egorin, Andrey; Tokar, Eduard; Tutov, Mikhail; Marinin, Dmitry; Avramenko, Valentin

    2017-01-05

    The origin of the emergence of radioactive contamination not removable in the process of acid-base regeneration of ion-exchange resins used in treatment of technological media and liquid radioactive waste streams has been determined. It has been shown that a majority of cesium radionuclides not removable by regeneration are bound to inorganic deposits on the surface and inside the ion-exchange resin beads. The nature of the above inorganic inclusions has been investigated by means of the methods of electron microscopy, IR spectrometry and X-ray diffraction. The method of decontamination of spent ion-exchange resins and zeolites contaminated with cesium radionuclides employing selective resorcinol-formaldehyde resins has been suggested. Good prospects of such an approach in deep decontamination of spent ion exchangers have been demonstrated. Copyright © 2016 Elsevier B.V. All rights reserved.

  2. Spent fuel transport in fuel cycle

    International Nuclear Information System (INIS)

    Labrousse, M.

    1977-01-01

    The transport of radioactive substances is a minor part of the fuel cycle because the quantities of matter involved are very small. However the length and complexity of the cycle, the weight of the packing, the respective distances between stations, enrichment plants and reprocessing plants are such that the problem is not negligible. In addition these transports have considerable psychological importance. The most interesting is spent fuel transport which requires exceptionally efficient packaging, especially where thermal and mechanical resistance are concerned. To meet the safety criteria necessary for the protection of both public and users it was decided to use the maximum capacity consistent with rail transport and to avoid coolant fluids under pressure. Since no single type of packing is suitable for all existing stations an effort has been made to standardise handling accessories, and future trands are towards maximum automation. A discussion on the various technical solutions available for the construction of these packing systems is followed by a description of those used for the two types of packaging ordered by COGEMA [fr

  3. Dehydrating process experiment on spent ion-exchange resin sludge by Funda Filter

    International Nuclear Information System (INIS)

    Hasegawa, Tatsuo; Ishino, Kazuyuki

    1977-01-01

    In nuclear power plants, Funda Filters are employed to dehydrate spent powdery ion-exchange resin sludge. The Funda Filter is very effective for eliminating small rust components contained in spent powdery resin slurry; however, in the drying process, the complete drying of spent powdery resin is very difficult because the filter cake of resin on the horizontal filter leaf is likely to crack and let out steam and hot air through the cracks. This paper deals with the results of experiments conducted to clarify the detailed phenomena of dehydration so the above problem could be solved. The above experiments were made on the precoating and drying of granular ion-exchange resin slurry that had not yet been put to practical use. The experiments were composed of one fundamental and one operational stage. In the fundamental experiment, the dehydration properties and dehydration mechanism of resins were made clear, and the most effective operational method was established through the operational experiments conducted using large-scale Funda Filter test equipment under various conditions. (auth.)

  4. Formulation study on immobilization of spent ion exchange resins in polymer cement

    International Nuclear Information System (INIS)

    Xia Lili; Lin Meiqiong; Bao Liangjin; Fan Xianhua

    2006-01-01

    The aim of this study is to develop a formulation of cement-solidified spent radioactive ion exchange resin form. The solidified form consists of a sort of composite cement, epoxide resin emulsion, and spent ion exchange resins. The composite cement is made up of quick-setting sulphoaluminate cement, silica powder, zeolite, and fly ash in the proportion 1:0.05:0.10:0.05. Sixteen combinations of composite cement, epoxide resin emulsion and mixed anion-cation exchange resins are selected according to a three-factors-four-levels normal design table with the compression strength as the evaluation criterion. The resulted formulation is as follows: the mass ratio of polymer emulsion to composite cement is 0.55:1, the loading of mixed anion-cation exchange resins is 0.3, and the anionic-to-cationic exchange resins ratio is 2:1. The polymer cement solidified forms were tested after 28 d curing for Cs + and Sr 2+ leaching rates, pH and conductivity of the leaching water, and radiation-resistant property in addition to their compressive strength. The measurement results indicate that the performance of thus prepared solidified forms can meet the requirements of the National Standard GB14569.1-93 for near earth's surface disposal of low radioactive waste. (authors)

  5. Sampling and characterization of spent exchange resins of Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Varani, Jose L.; Cernadas, D.; Iglesias, Alberto M.; Raffo Calderon, Maria del C.

    2004-01-01

    The present storage facilities for spent resins in Atucha I NPP would be full within the next 5 years, accordingly some tasks are being planned to conditioning these waste and so generate enough storage capacity for the remaining life of the plant. Among this tasks is the characterization of spent resins that has several objectives: to know their chemical and radiochemical composition; the homogeneity of these parameters in the total volume of spent resins; the existence or not of compact zones; the proportion of 'crud'; the breakage degree of the micro-spheres; etc. The first step was to analyse the criterion to follow for sampling resins in the storage deposit of 40 m 3 . In order to take some samples from different points, a special device was required. It had to be introduced closed in the resin bed, opened to take the sample and then closed again to return to the surface. A device used in cereal industry for sampling silos to different depths was modified in its internal capacity for reducing operator dose and increasing the length of rod in order to reaching the bottom of the pit. The device was tested in cold mock up before to taking actual samples. Active resins samples, five in total up to now, were taken from deposit to different depths and kept in lead containers. After analysing the samples, the following average results were extracted: 1.7 x 10 5 Bq/g of Co-60, 9.7 x 10 5 Bq/g of Cs-137 and 774 Bq/g of total alpha, which corresponds to intermediate activity waste. The differences between the values of activity of the different samples are of up to 310 % for Co-60 and of up to 788 % for the Cs-137 what indicates a great inhomogeneity. The direct observation of resin grains, placed in a transparent glass burette, did not demonstrate an important proportion of broken or divided resins. (author)

  6. Leaching of Co and Cs from spent ion exchange resins in cement ...

    Indian Academy of Sciences (India)

    Unknown

    2003-08-22

    Aug 22, 2003 ... matrix with a waste load of 290–350 (kg/m3) spent cation exchange resins, was measured for 60Co : (1,1–4,0) ×. 10–6 (cm2/d) and ... active waste management for a future Serbian radioactive waste disposal centre. Keywords. ... The cement specimens were prepared from construction cement which is ...

  7. Studies on the incorporation of spent ion exchange resins from nuclear power plants into bitumen and cement

    International Nuclear Information System (INIS)

    Bonnevie-Svendsen, M.; Tallberg, K.; Aittola, P.; Tollbaeck, H.

    1976-01-01

    The joint Nordic incorporation experiments should provide technical data needed for the assessment of solidification techniques for wastes from nuclear reactors in the Nordic countries. Spent ion exchange resins are a main fraction of such wastes, and more knowledge about their incorporation is wanted. The effects of simulated and real ion exchange wastes on the quality of bitumen and cement incorporation products were studied. Blown and distilled bitumen and three Portland cement qualities were used. Product characterizations were based on properties relevant for safe waste management, storage, transport and disposal. The applicability and relevance of established and suggested tests is discussed. Up to 40-60% dry resin could be incorporated into bitumen without impairing product qualities. Products with higher resin contents were found to swell in contact with water. The products had a high leach resistance. Their form stability was improved by incorporated resins. Product qualities appeared to be less affected by physico-chemical variables than by mechanical process parameters. Pure resin-cement products tend to decompose in water. Product qualities were strongly affected by a variety of physico-chemical process parameters, and integer products were only obtained within narrow tolerance limits. Caesium was rapidly leached out. To attain integer products and improved leach resistance within technically acceptable tolerance limits it was necessary to utilize stabilizing and caesium-retaining additives such as Silix and vermiculite. Under the present conditions the water content of the resins limited the amounts that could be incorporated in 40-50wt% or about 70vol.% water-saturated (containing 20-40% dry) resin. (author)

  8. Wet oxidative destruction of spent ion-exchange resins using hydrogen peroxide

    International Nuclear Information System (INIS)

    Srinivas, C.; Ramaswamy, M.; Theyyunni, T.K.

    1994-01-01

    Spent organic ion exchange resins are generated in large quantities during the operation of nuclear facilities. Wet oxidation as a mode of treatment of these gel-type ion exchange resins was investigated using H 2 O 2 as oxidant in the presence of CuSO 4 as catalyst. Experiments using commercial samples were conducted at 95-100 deg C under reflux conditions at atmospheric pressure. It was found that the reaction of cation resin with H 2 O 2 was instantaneous whereas with anion resin, there was a lag time. For efficient utilization of the oxidant, low rate of addition of H 2 O 2 , 0.01M concentration of CuSO 4 and neutral pH in mixed resin reactions, were found to be useful. Foaming was noted during reactions involving anion resins. This could be controlled by silicone-based agents. The residual solution left after resin oxidation is aqueous in nature and is expected to contain all the radioactivity originally present in the resin. Preliminary experiments showed that it could be efficiently trapped using available inorganic sorbents. Wet oxidation system offers a simple method of converting organic waste into environmentally acceptable inorganic products. (author). 8 refs., 6 figs., 2 tabs

  9. Transporting spent reactor fuel: allegations and responses

    International Nuclear Information System (INIS)

    Jefferson, R.M.

    1983-03-01

    A January 1982 monthly newsletter from the Council on Economic Priorities (CEP) was entirely devoted to the presentation of a broad-ranging series of allegations that the transportation of spent fuel in particular, and other high-level radioactive materials by inference is currently being conducted in this country in an unsafe manner. This newsletter preceded the release of a book authored by Marvin Resnikoff on the same subject by over a year. This book titled The Next Nuclear Gamble contained substantially the same allegations as the newsletter, although the book devoted space to a greatly increased number of specific examples. This paper reduces those allegations contained in the executive summary and the recommendations contained in the last chapter of the book to a manageable number by combining the many specific issues into a few topics. Each of these topics is then addressed. As such, this is an abbreviated analysis of The Next Nuclear Gamble and does not address much of the fine detail. In spite of that, it would be possible to address each of the details within the book on a similar basis. The intent of this document is to provide background information for those who are questioned on the validity of the allegations made by the CEp

  10. Technical assessment of NOx generation from vitrification process of spent ion-exchange resin

    International Nuclear Information System (INIS)

    Park, S. C.; Yang, K. H.; Kim, H. S.; Lu, C. S.; Lee, K. H.; Hwnag, T. W.; Shin, S. W.

    2001-01-01

    When the radioactive spent ion-exchange resin is being treated in vitrification system, due to the nitrogen in the anion exchange resin media and the nitrogen in air inleaked to the system, the nitrogen oxide (NOx) is generated from both glass melter and the second combustion chamber among the unit-processes in the vitrification plant. The NOx is very hazardous to environment and to human health the emission limit of NOx is regulated very severely. In this study, the NOx generation characteristics are technically analyzed based on the demonstration-test resultes conducted recently by burning simulated spent resin. When burning 30kg/h of simulated resin in CCM under 50% of excess the theoretically needed, the NOx was measured as between 3000 ∼ 3500ppm after 1h of transient test period. And when only the propane is burning in PCC without resin burning in CCM, the concentration of NOx exceeded the detectable limit(4000 ppm) of PGA. The former and the latter were considered as the fuel NOx and the thermal NOx respectively

  11. Spent fuel transportation accident: a state's involvement

    International Nuclear Information System (INIS)

    Neuweg, M.

    1978-01-01

    On February 9, 1978 at 8:20 p.m., the duty officer for the Illinois Radiological Assistance Team was notified that a shipment containing uranium and plutonium was involved in an accident near Gibson City, Illinois on Route 54. It was reported that a pig containing an unknown amount of uranium and plutonium was involved. The Illinois District 6A State Police were called to the scene and secured the area. The duty officer in the meantime learned after numerous telephone calls, approximately 1 hour after the first notice was received, that the pig actually was a 48,000 pound cask containing 6 spent fuel rods and the tractor-trailer had split apart and was blocking one lane of the highway. The shipment had departed from Dresden Nuclear Power Station, Morris, Illinois, enroute to Babcox and Wilcox in Lynchburg, Virginia. Initial reports indicated the vehicle had split apart. Actually, the semi-trailer bed had buckled beneath the cask due to apparent excess stress. The cask remained entirely intact and was not damaged, but the state highway was closed to traffic. The State Radiological Assistance Team was dispatched and arrived on the scene at 12:45 a.m. Immediate radiation monitoring revealed a reading of 4 milliroentgen per hour at 10 feet from the cask. No contamination existed nor was anyone exposed to radiation unnecessarily. The cask was transferred to a Tri-State semi-trailer vehicle the following morning at approximately 6:30 a.m. At 9:30 a.m., February 10, the new vehicle was again enroute to its destination. This incident demonstrated typical occurrences involving transportation radiation accident: misinformation and/or lack of information on the initial response notification, inaccuracies of radiation monitorings at the scene of the accident, inconsistencies concerning the occurrences of the accident and unfamiliar terminology utilized by personnel first on the scene, i.e., pig, cask, vehicle split apart, etc

  12. Comparison of the Transportation Risks Resulting from Accidents during the Transportation of the Spent Fuel

    International Nuclear Information System (INIS)

    Jeong Jong Tae; Cho, Dong Kuen; Choi, Heui Joo; Choi, Jong Won

    2007-01-01

    The safe, environmentally sound and publicly acceptable disposal of high level wastes and spent fuels is becoming a very important issue. The operational safety assessment of a repository including a transportation safety assessment is a fundamental part in order to achieve this goal. According to the long term management strategy for spent fuels in Korea, they will be transported from the spent fuel pools in each nuclear power plant to the central interim storage facility (CISF) which is to start operation in 2016. Therefore, we have to determine the safe and economical logistics for the transportation of these spent fuels by considering their transportation risks and costs. In this study, we developed four transportation scenarios by considering the type of transportation casks and transport means in order to suggest safe and economical transportation logistics for spent fuels. Also, we estimated and compared the transportation risks resulting from the accidents during the transportation of spent fuels for these four transportation scenarios

  13. Spent Fuel Transportation Package Performance Study - Experimental Design Challenges

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  14. Development of nuclear spent fuel Maritime transportation scenario

    International Nuclear Information System (INIS)

    Yoo, Min; Kang, Hyun Gook

    2014-01-01

    Spent fuel transportation of South Korea is to be conducted through near sea because it is able to ship a large amount of the spent fuel far from the public comparing to overland transportation. The maritime transportation is expected to be increased and its risk has to be assessed. For the risk assessment, this study utilizes the probabilistic safety assessment (PSA) method and the notions of the combined event. Risk assessment of maritime transportation of spent fuel is not well developed in comparison with overland transportation. For the assessment, first, the transportation scenario should be developed and categorized. Categories are assorted into the locations, release aspects and exposure aspects. This study deals with accident that happens on voyage and concentrated on ship-ship collision. The collision accident scenario is generated with event tree analysis. The scenario will be exploited for the maritime transportation risk model which includes consequence and accident probability

  15. Incineration of spent ion exchange resins in a triphasic mixture at Belgoprocess

    International Nuclear Information System (INIS)

    Deckers, J.; Luycx, P.

    2003-01-01

    Up to 1998, spent ion exchange resins have been fed to the incinerator in combination with various other solid combustible wastes at Belgoprocess. However, thanks to sustained efforts to reduce radioactive waste production in all nuclear facilities in Belgium, the annual production of solid combustible waste is now much too small to allow this practice to be continued. Since the incinerator at Belgoprocess is not capable of treating spent ion exchange resins as such, it was decided to adopt the use of foam as a carrier to feed the resins to the incinerator. The mixture is a pseudohomogeneous charged foam, ensuring easy handling and allowing incineration in the existing furance, while a number of additives may be included, such as oil to increase the calorific value of the mixture and accelerate combustion. The first incineration campaign of spent ion exchange resins in a triphasic foam mixture, in conjunction with other liquid and solid combustible wastes, will be started in January 2000. The foam, comprising 70% by weight of resins, 29% by weight of water and 1% by weight of surfactant will be pulverized in the incinerator through an injection lance, at a feed rate of 40 to 100 kg/h. The incinerator and associated off-gas treatment system can be operated at standard conditions. Belgoprocess is the subsidiary of the Belgian national agency for the management of radioactive waste, known by its Dutch and French acronyms, NIRAS and ONDRAF respectively. The company ensures the treatment, conditioning and interim storage of nearly all radioactive waste produced in Belgium. (orig.)

  16. Quantum-CEP processing spent ion exchange resins from nuclear power stations

    International Nuclear Information System (INIS)

    Kaczmarsky, Myron

    1997-01-01

    Quantum-CEP (Q-CEP) is an innovative and proprietary technology developed by Molten Metal Technology, Inc, which can process radioactive and mixed waste streams to decontaminate and recover resources of commercial value while achieving significant volume reduction and radionuclide stabilization. A Q-CEP facility, located in Oak Ridge, Tennessee, processes low-level radioactive spent ion exchange resins (IER) from U.S commercial nuclear power plants. The first campaign processing low level radioactive spent IER was successfully completed in December 1996. Other milestones, since December, include operating parallel Trains A and B simultaneously and processing 25,000 lb. dry resin (50,000 lb. wet resin) or six equivalent High Integrity Containers (HICs) in one batch campaign, in March; and processing 50,000 lb. dry resin or 12 equivalent HICs in one batch campaign in May. This paper presents results from the March campaign (97-008) in which 25,000 lb. of dry spent IER from five nuclear power stations were processed. This campaign has been selected since it is representative of campaigns completed during the first five months of operation. Key highlights for this campaign include processing six HICs in batch campaign while achieving a volume reduction of 24: 1. Key performance targets for the facility are to process an average of six HICs per campaign batch and achieve a volume reduction of 30: 1. The average batch size and other performance parameters have steadily improved during the initial operating period with radioactive resin. The progress was dramatically demonstrated by the May campaign during which 12 HICs were processed - achieving a volume reduction estimated to exceed 50: 1. The campaigns in March and May demonstrate that the facility's design and technology are capable of achieving and even exceeding the facility's key target performance parameters

  17. Spent fuel storage and transportation - ANSTO experience

    International Nuclear Information System (INIS)

    Irwin, Tony

    2002-01-01

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

  18. Transportation capabilities study of DOE-owned spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-10-01

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

  19. Engineering study for the treatment of spent ion exchange resin resulting from nuclear process applications

    International Nuclear Information System (INIS)

    Place, B.G.

    1990-09-01

    This document is an engineering study of spent ion exchange resin treatment processes with the purpose of identifying one or more suitable treatment technologies. Classifications of waste considered include all classes of low-level waste (LLW), mixed LLW, transuranic (TRU) waste, and mixed TRU waste. A total of 29 process alternatives have been evaluated. Evaluation parameters have included economic parameters (both total life-cycle costs and capital costs), demonstrated operability, environmental permitting, operational availability, waste volume reduction, programmatic consistency, and multiple utilization. The results of this study suggest that there are a number of alternative process configurations that are suitable for the treatment of spent ion exchange resin. The determinative evaluation parameters were economic variables (total life-cycle cost or capital cost) and waste volume reduction. Immobilization processes are generally poor in volume reduction. Thermal volume reduction processes tend to have high capital costs. There are immobilization processes and thermal volume reduction processes that can treat all classifications of spent ion exchange resin likely to be encountered. 40 refs., 19 figs., 17 tabs

  20. Decontamination of spent ion-exchangers contaminated with cesium radionuclides using resorcinol-formaldehyde resins

    Energy Technology Data Exchange (ETDEWEB)

    Palamarchuk, Marina, E-mail: marina_p@ich.dvo.ru; Egorin, Andrey; Tokar, Eduard; Tutov, Mikhail; Marinin, Dmitry; Avramenko, Valentin

    2017-01-05

    Highlights: • Cesium radionuclides not removable by regeneration are bound to silicate deposits. • Application of RFR substantially increases cesium desorption from an ion-exchanger. • The radwaste volume was reduced at least 2-fold for zeolites and 10-fold for SIER. • The distribution coefficient values for RFR were high (K{sub d} > 10{sup 4}) after 6 regenerations. • The volume of secondary waste formed after regeneration of RFR was reduced 600-fold. - Abstract: The origin of the emergence of radioactive contamination not removable in the process of acid-base regeneration of ion-exchange resins used in treatment of technological media and liquid radioactive waste streams has been determined. It has been shown that a majority of cesium radionuclides not removable by regeneration are bound to inorganic deposits on the surface and inside the ion-exchange resin beads. The nature of the above inorganic inclusions has been investigated by means of the methods of electron microscopy, IR spectrometry and X-ray diffraction. The method of decontamination of spent ion-exchange resins and zeolites contaminated with cesium radionuclides employing selective resorcinol-formaldehyde resins has been suggested. Good prospects of such an approach in deep decontamination of spent ion exchangers have been demonstrated.

  1. Decontamination of spent ion-exchangers contaminated with cesium radionuclides using resorcinol-formaldehyde resins

    International Nuclear Information System (INIS)

    Palamarchuk, Marina; Egorin, Andrey; Tokar, Eduard; Tutov, Mikhail; Marinin, Dmitry; Avramenko, Valentin

    2017-01-01

    Highlights: • Cesium radionuclides not removable by regeneration are bound to silicate deposits. • Application of RFR substantially increases cesium desorption from an ion-exchanger. • The radwaste volume was reduced at least 2-fold for zeolites and 10-fold for SIER. • The distribution coefficient values for RFR were high (K d > 10 4 ) after 6 regenerations. • The volume of secondary waste formed after regeneration of RFR was reduced 600-fold. - Abstract: The origin of the emergence of radioactive contamination not removable in the process of acid-base regeneration of ion-exchange resins used in treatment of technological media and liquid radioactive waste streams has been determined. It has been shown that a majority of cesium radionuclides not removable by regeneration are bound to inorganic deposits on the surface and inside the ion-exchange resin beads. The nature of the above inorganic inclusions has been investigated by means of the methods of electron microscopy, IR spectrometry and X-ray diffraction. The method of decontamination of spent ion-exchange resins and zeolites contaminated with cesium radionuclides employing selective resorcinol-formaldehyde resins has been suggested. Good prospects of such an approach in deep decontamination of spent ion exchangers have been demonstrated.

  2. Remarks on the transportation of spent fuel elements

    International Nuclear Information System (INIS)

    Krull, W.

    1986-01-01

    In this chapter topics discussed are the need for contracts, a transport company and risk insurance. Also, a section on transportation covers cranes, subpressure, contamination, cask limitations, physical protection and shipping. Reprocessing discusses minimum reprocessing batch and spent fuel. Finally, economical considerations concerning transportation and reprocessing are given

  3. Arrival condition of spent fuel after storage, handling, and transportation

    International Nuclear Information System (INIS)

    Bailey, W.J.; Pankaskie, P.J.; Langstaff, D.C.; Gilbert, E.R.; Rising, K.H.; Schreiber, R.E.

    1982-11-01

    This report presents the results of a study conducted to determine the probable arrival condition of spent light-water reactor (LWR) fuel after handling and interim storage in spent fuel storage pools and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits for storing, handling, and transporting unconsolidated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed. 95 references, 3 figures, 17 tables

  4. Transport of HLW and spent fuel in Japan

    International Nuclear Information System (INIS)

    Nakajima, Mitsuo.

    1995-01-01

    In Japan, transport of HLW started in spring, 1995 when vitrified waste was returned to Japan from Europe. Domestic transport of spent fuel has been carried out since 1978 from domestic nuclear power plant to the reprocessing plant in Tokai-mura, Ibaraki Prefecture, amounting to some 720 tons by the end of 1994. A commercial reprocessing plant is under construction in Rokkasho-mura, Aomori Prefecture and the transport thereto of high burn-up spent fuel is scheduled to start in 1997. A 150-ton wharf crane and a dedicated transport vehicle were completed in 1994 for unloading and overland transport of HLW casks, and specifically designed casks and an exclusive-use vessel are being prepared for domestic transport of high burn-up spent fuel

  5. Potential exposures and health effects from spent fuel transportation

    International Nuclear Information System (INIS)

    Rogers, V.C.; Sandquist, G.M.; Sutherland, A.A.

    1987-01-01

    The radiation exposures and consequent health effects associated with normal operations and postulated accidents during transportation of spent fuel have been analyzed and evaluated and the results have been summarized in the Final Environmental Assessments issued by DOE. For normal, accident-free transport of spent fuel, radiation exposures arise from both gamma and neutron sources within the spent fuel cask. The neutrons result from the spontaneous fission of transuranic nuclides in the spent fuel. The neutron dose flux component was modeled using DISNEL, a generalized, one-dimensional, multi-energy group neutronics code. Computer program PATHRAE-T was then developed from the EPA code PATHRAE and was employed to determine the total, combined dose field, including both ground and sky scatter of neutrons and gamma photons for any position around a truck or rail spent fuel cask. Four activity classes, viz., caravan, traffic obstruction, resident and pedestrian proximity, and servicing of the cask transport vehicle were reviewed for maximum individual exposure assessments. Projected doses for typical activities under maximum exposure conditions were 6 mrem or less per event. A spent fuel rail cask containing up to 14 PWR spent fuel assemblies could conceivably be involved in a variety of rail related transportation accidents. PATHRAE-T was used to estimate doses from rail cask accidents involving the release of radioactive nuclides although a release from such accidents is highly unlikely. The maximum individual exposure, primarily due to inhalation, is about 10 rem and occurs about 70 meters downwind

  6. Integrated risk assessment for spent fuel transportation using developed software

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Mi Rae; Christian, Robby; Kim, Bo Gyung; Almomani, Belal; Ham, Jae Hyun; Kang, Gook Hyun [KAIST, Daejeon (Korea, Republic of); Lee, Sang hoon [Keimyung University, Daegu (Korea, Republic of)

    2016-05-15

    As on-site spent fuel storage meets limitation of their capacity, spent fuel need to be transported to other place. In this research, risk of two ways of transportation method, maritime transportation and on-site transportation, and interim storage facility were analyzed. Easier and integrated risk assessment for spent fuel transportation will be possible by applying this software. Risk assessment for spent fuel transportation has not been researched and this work showed a case for analysis. By using this analysis method and developed software, regulators can get some insights for spent fuel transportation. For example, they can restrict specific region for preventing ocean accident and also they can arrange spend fuel in interim storage facility avoiding most risky region which have high risk from aircraft engine shaft. Finally, they can apply soft material on the floor for specific stage for on-site transportation. In this software, because we targeted Korea, we need to use Korean reference data. However, there were few Korean reference data. Especially, there was no food chain data for Korean ocean. In MARINRAD, they used steady state food chain model, but it is far from reality. Therefore, to get Korean realistic reference data, dynamic food chain model for Korean ocean need to be developed.

  7. Integrated risk assessment for spent fuel transportation using developed software

    International Nuclear Information System (INIS)

    Yun, Mi Rae; Christian, Robby; Kim, Bo Gyung; Almomani, Belal; Ham, Jae Hyun; Kang, Gook Hyun; Lee, Sang hoon

    2016-01-01

    As on-site spent fuel storage meets limitation of their capacity, spent fuel need to be transported to other place. In this research, risk of two ways of transportation method, maritime transportation and on-site transportation, and interim storage facility were analyzed. Easier and integrated risk assessment for spent fuel transportation will be possible by applying this software. Risk assessment for spent fuel transportation has not been researched and this work showed a case for analysis. By using this analysis method and developed software, regulators can get some insights for spent fuel transportation. For example, they can restrict specific region for preventing ocean accident and also they can arrange spend fuel in interim storage facility avoiding most risky region which have high risk from aircraft engine shaft. Finally, they can apply soft material on the floor for specific stage for on-site transportation. In this software, because we targeted Korea, we need to use Korean reference data. However, there were few Korean reference data. Especially, there was no food chain data for Korean ocean. In MARINRAD, they used steady state food chain model, but it is far from reality. Therefore, to get Korean realistic reference data, dynamic food chain model for Korean ocean need to be developed

  8. Study of plasma off-gas treatment from spent ion exchange resin pyrolysis.

    Science.gov (United States)

    Castro, Hernán Ariel; Luca, Vittorio; Bianchi, Hugo Luis

    2017-03-23

    Polystyrene divinylbenzene-based ion exchange resins are employed extensively within nuclear power plants (NPPs) and research reactors for purification and chemical control of the cooling water system. To maintain the highest possible water quality, the resins are regularly replaced as they become contaminated with a range of isotopes derived from compromised fuel elements as well as corrosion and activation products including 14 C, 60 Co, 90 Sr, 129 I, and 137 Cs. Such spent resins constitute a major proportion (in volume terms) of the solid radioactive waste generated by the nuclear industry. Several treatment and conditioning techniques have been developed with a view toward reducing the spent resin volume and generating a stable waste product suitable for long-term storage and disposal. Between them, pyrolysis emerges as an attractive option. Previous work of our group suggests that the pyrolysis treatment of the resins at low temperatures between 300 and 350 °C resulted in a stable waste product with a significant volume reduction (>50%) and characteristics suitable for long-term storage and/or disposal. However, another important issue to take into account is the complexity of the off-gas generated during the process and the different technical alternatives for its conditioning. Ongoing work addresses the characterization of the ion exchange resin treatment's off-gas. Additionally, the application of plasma technology for the treatment of the off-gas current was studied as an alternative to more conventional processes utilizing oil- or gas-fired post-combustion chambers operating at temperatures in excess of 1000 °C. A laboratory-scale flow reactor, using inductively coupled plasma, operating under sub-atmospheric conditions was developed. Fundamental experiments using model compounds have been performed, demonstrating a high destruction and removal ratio (>99.99%) for different reaction media, at low reactor temperatures and moderate power consumption

  9. Prospect for spent fuel transportation system in China

    International Nuclear Information System (INIS)

    Li, X.Q.; Jiang, Y.Q.

    1998-01-01

    With the arising of spent fuel from nuclear power plants, a national policy of a closed nuclear fuel cycle has been determined. Following being stored at reactor sites for at least 5 years (with storage maximum of 10 years), spent fuel will be transferred into an away-from-reactor wet centralized storage facility, waiting for reprocessing. Therefore, China's spent fuel management activities would be involved in at-reactor storage, transportation, away-from-reactor storage and reprocessing, as well as certainly radioactive waste management. The status of the nuclear fuel cycle will be briefly discussed and the potential spent fuel transportation system in China is discussed in detail, covering administration, flask, terminals and transportation by sea, rail and road. (authors)

  10. Transport of oxide spent fuel. Industrial experience of COGEMA

    International Nuclear Information System (INIS)

    Lenail, B.

    1983-01-01

    COGEMA is ruling all transports of spent fuel to La Hague reprocessing plant. The paper summarizes some aspects of the experience gained in this field (road, rail and sea transports) and describes the standards defined by COGEMA as regards transport casks. These standards are as follows: - casks of dry type, - casks of the maximum size compatible with rail transports, - capability to be unloaded with standardized equipment and following standard procedures

  11. Shielding Performance Measurements of Spent Fuel Transportation Container

    Directory of Open Access Journals (Sweden)

    SUN Hong-chao

    2015-11-01

    Full Text Available The safety supervision of radioactive material transportation package has been further stressed and implemented. The shielding performance measurements of spent fuel transport container is the important content of supervision. However, some of the problems and difficulties reflected in practice need to be solved, such as the neutron dose rate on the surface of package is too difficult to measure exactly, the monitoring results are not always reliable, etc. The monitoring results using different spectrometers were compared and the simulation results of MCNP runs were considered. An improvement was provided to the shielding performance measurements technique and management of spent fuel transport.

  12. Risk of transporting spent nuclear fuel by train

    International Nuclear Information System (INIS)

    Elder, H.K.

    1981-12-01

    This paper presents results of a study which analyzes the risk of transporting spent fuel by train. The risk assessment methodology consists of 4 basic steps: (1) a description of the system being analyzed; (2) identification of sequences of events that could lead to a release of material during transportation; (3) evaluation of the probability and consequences of each release sequence; and (4) assessment of the risk and evaluation of the results. The conclusion reached was that considering the substantial benefits derived from the fuel, the current spent fuel transportation system poses reasonably low risks

  13. Comparison of the transportation risks for the spent fuel in Korea for different transportation scenarios

    International Nuclear Information System (INIS)

    Jeong, Jongtae; Cho, D.K.; Choi, H.J.; Choi, J.W.

    2011-01-01

    According to the long term management strategy for spent fuels in Korea, they will be transported from the spent fuel pools in each nuclear power plant to the central interim storage facility (CISF) which is to start operation in 2016. At the start of the operation of the final repository (FR), by the year 2065, transport will then take place between the CISF and the FR. Therefore, we have to determine the safe and economical logistics for the transportation of these spent fuels by considering their transportation risks and costs. In this study, we developed four transportation scenarios for a maritime transportation by considering the type of transportation casks and transport means in order to suggest safe and economical transportation logistics for the spent fuels in Korea. And, we estimated and compared the transportation risks for these four transportation scenarios. Also, we estimated and compared the transportation risks resulting from accidents during the transportation of PWR and PHWR spent fuels by road trailers from the CISF and the FR. From the results of this study, we found that risks resulting from accidents during the transportation of the spent fuels have a very low radiological risk activity with a manageable safety and health consequences. The results of this study can be used as basic data for the development of safe and economical logistics for a transportation of the spent fuels in Korea by considering the transportation costs for the four scenarios which will be needed in the near future.

  14. Accurate determination of 41Ca concentrations in spent resins from the nuclear industry by Accelerator Mass Spectrometry

    International Nuclear Information System (INIS)

    Nottoli, Emmanuelle; Bourlès, Didier; Bienvenu, Philippe; Labet, Alexandre; Arnold, Maurice; Bertaux, Maité

    2013-01-01

    The radiological characterisation of nuclear waste is essential for managing storage sites. Determining the concentration of Long‐Lived RadioNuclides (LLRN) is fundamental for their long-term management. This paper focuses on the measurement of low 41 Ca concentrations in ions exchange resins used for primary fluid purification in Pressurised Water Reactors (PWR). 41 Ca concentrations were successfully measured by Accelerator Mass Spectrometry (AMS) after the acid digestion of resin samples, followed by radioactive decontamination and isobaric suppression through successive hydroxide, carbonate, nitrate and final CaF 2 precipitations. Measured 41 Ca concentrations ranged from 0.02 to 0.03 ng/g, i.e. from 0.06 to 0.09 Bq/g. The 41 Ca/ 60 Co activity ratios obtained were remarkably reproducible and in good agreement with the current ratio used for resins management. - Highlights: • In the context of radioactive waste management, this study aimed at measuring 41 Ca in spent resins using Accelerator Mass Spectrometry. • A chemical treatment procedure was developed to quantitatively recover calcium in solution and selectively extract it. • Developed firstly on synthetic matrices, the chemical treatment procedure was then successfully applied to real resin samples. • Accelerator mass spectrometry allowed measuring concentrations of 41 Ca in spent resins as low as 0.02 ng/g of dry resin. • Final results are in agreement with current data used for spent resins management

  15. Mixture Design and Its Application in Cement Solidification for Spent Resin

    International Nuclear Information System (INIS)

    Gan, Xueying; Lin, Meiqing; Chen, Hui

    1994-01-01

    The study is aimed to assess the usefulness of the mixture design for spent resin immobilization in cement. Although a considerable amount of research has been carried out to determine the limits for the composition of an acceptable resin-cement mixture, no efficient experimental strategy exists that explores the full properties of waste form against composition relationship. In order to gain an overall view, this report introduces the method of mixture design and mixture analysis, and describes the design of experiment of the 5-component mixture with the constraint conditions. The mathematic models of 28-day compressive strength varying with the ingredients are fitted, and the main effect and interaction effect of two ingredients are identified quantitatively along with the graphical interpretation using the response trace plot and contour plots

  16. MTR spent fuel transport and handling experience

    Energy Technology Data Exchange (ETDEWEB)

    Roland, Vincent [TRANSNUCLEAIRE (France)

    1999-07-01

    The present paper describes the last MTR transport operations performed by TN in exotic countries, as well as within Europe. Each transport is specific and must be very carefully prepared, because all MTR fuels are generally very specific to each research reactor. Their characteristics (i.e. type, dimensions, irradiation...) have to be precisely identified because, for instance, they are not always well-known due to their period of storage. We will mainly talk about the International Shipments. (author)

  17. Spent fuels transportation coming from Australia; Transport de combustible use en provenance d'Australie

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

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

  18. Characterization of radioactive wastes - spent ion-exchange resins and charcoal filter beds

    International Nuclear Information System (INIS)

    Silva, Rozilene Elaine; Isiki, Vera Lucia Keiko; Goes, Marcos Maciel de; Potiens Junior, Ademar Jose; Dellamano, Jose Claudio; Vicente, Roberto

    2009-01-01

    In the present paper we report the initial results of the work undertaken at the Radioactive Waste Management Laboratory (RWML), in Sao Paulo, Brazil, to develop sampling procedures and analytical methods applied to the characterization of radioactive wastes, specifically spent ion-exchange resins and charcoal filter beds generated at the IEA-R1 Research Reactor operated by the Nuclear Energy Research Institute. The work objectives are to characterize those wastes to comply with regulatory requirements, to generate data to support the development of treatment processes, and to improve characterization methods and laboratorial infrastructure. (author)

  19. Solidification of radioactive liquid wastes. A comparison of treatment options for spent resins and concentrates

    International Nuclear Information System (INIS)

    Roth, A.; Willmann, F.; Ebata, M.; Wendt, S.

    2008-01-01

    Ion exchange is one of the most common and effective treatment methods for liquid radioactive waste. However, spent ion exchange resins are considered to be problematic waste that in many cases require special approaches and pre-conditioning during its immobilization to meet the acceptance criteria for disposal. Because of the function that they fulfill, spent ion exchange resins often contain high concentrations of radioactivity and pose special handling and treatment problems. Another very common method of liquid radioactive waste treatment and water cleaning is the evaporation or diaphragm filtration. Both treatment options offer a high volume reduction of the total volume of liquids treated but generate concentrates which contain high concentrations of radioactivity. Both mentioned waste streams, spent resins as well as concentrates, resulting from first step liquid radioactive waste treatment systems have to be conditioned in a suitable manner to achieve stable waste products for final disposal. The most common method of treatment of such waste streams is the solidification in a solid matrix with additional inactive material like cement, polymer etc. In the past good results have been achieved and the high concentration of radioactivity can be reduced by adding the inactive material. On the other hand, under the environment of limited space for interim storage and the absence of a final repository site, the built-up of additional volume has to be considered as very critical. Moreover, corrosive effects on cemented drums during long-term interim storage at the surface have raised doubts about the long-term stability of such waste products. In order to avoid such disadvantages solidification methods have been improved in order to get a well-defined product with a better load factor of wastes in the matrix. In a complete different approach, other technologies solidify the liquid radioactive wastes without adding of any inactive material by means of drying

  20. An Indian perspective for transportation and storage of spent fuel

    International Nuclear Information System (INIS)

    Dey, P.K.

    2005-01-01

    The spent fuel discharged from the reactors are temporarily stored at the reactor pool. After a certain cooling time, the spent fuel is moved to the storage locations either on or off reactor site depending on the spent fuel management strategy. As India has opted for a closed fuel cycle for its nuclear energy development, reprocessing of the spent fuel, recycling of the reprocessed plutonium and uranium and disposal of the wastes from the reprocessing operations forms the spent fuel management strategy. Since the reprocessing operations are planned to match the nuclear energy programme, storage of the spent fuel in ponds are adopted prior to reprocessing. Transport of the spent fuel to the storage locations are carried out adhering to international and national guide lines. India is having 14 operating power reactors and three research reactors. The spent fuel from the two safeguarded BWRs are stored at-reactor (AR) storage pond. A separate wet storage facility away-from-reactor (AFR) has been designed, constructed and made operational since 1991 for additional fuel storage. Storage facilities are provided in ARs at other reactor locations to cater to 10 reactor-years of operation. A much lower capacity spent fuel storage is provided in reprocessing plants on the same lines of AR fuel storage design. Since the reprocessing operations are carried out on a need basis, to cater to the increased storage needs two new spent fuel storage facilities (SFSF) are being designed and constructed near the existing nuclear plant sites. India has mastered the technology for design, construction and operation of wet spent fuel storage facility meeting all the international standards Wet storage of the spent fuel is the most commonly adopted mode all over the world. Recently an alternate mode viz. dry storage has also been considered. India has designed, constructed and operated lead shielded dry storage casks and is operational at one site. A dry storage cask made of concrete

  1. BNFL's new spent fuel transport flask - Excellox 8

    International Nuclear Information System (INIS)

    McWilliam, D.S.

    2002-01-01

    Since British Nuclear Fuels plc (BNFL) was formed in 1971 its transport service has safely moved spent light water reactor fuel from many locations abroad to its fuel handling plants at Sellafield in the UK. To support this business a number of types of flasks have been designed and used. One of the types used has been the Excellox family of water-filled flasks. To support future business opportunities a new flask, designed to meet the requirements of the new IAEA transport regulations TS-R-1 (ST-1, Revised), has been developed. The flask will be a type B(U)F. This new flask design will maximise fuel carrying capacity to minimise transport costs. The design capacity of the new Excellox 8 flask is to be 12 pressurised water reactor or 32 boiling water reactor fuel assemblies. The objective of this BNFL project is to provide another economic spent nuclear fuel transport system, in support of BNFL transport business. (author)

  2. Probabilistic Risk Assessment on Maritime Spent Nuclear Fuel Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Christian, Robby; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    Spent nuclear fuel (SNF) management has been an indispensable issue in South Korea. Before a long term SNF solution is implemented, there exists the need to distribute the spent fuel pool storage loads. Transportation of SNF assemblies from populated pools to vacant ones may preferably be done through the maritime mode since all nuclear power plants in South Korea are located at coastal sites. To determine its feasibility, it is necessary to assess risks of the maritime SNF transportation. This work proposes a methodology to assess the risk arising from ship collisions during the transportation of SNF by sea. Its scope is limited to the damage probability of SNF packages given a collision event. The effect of transport parameters' variation to the package damage probability was investigated to obtain insights into possible ways to minimize risks. A reference vessel and transport cask are given in a case study to illustrate the methodology's application.

  3. Transportation of high-level waste and spent fuel

    International Nuclear Information System (INIS)

    Carlson, J.H.; Lake, W.H.; Thompson, J.H.

    1993-01-01

    The Office of Civilian Radioactive Waste Management (OCRWM) transportation program is a multifaceted undertaking to transport spent nuclear fuel from commercial reactors to temporary and permanent storage facilities commencing in 1998. One of the significant ingredients necessary to achieving this goal is the development and acquisition of shipping casks. Efforts to design and acquire high capacity casks is ongoing, as are efforts to purchase casks that can be made available using current technology. By designing casks that are optimized to the specifications of the older cooler spent fuel that will be shipped, and by designing to current NRC requirements, OCRWM's new generation of spent fuel casks will be more efficient and at least as safe as current cask designs. (J.P.N.)

  4. Solidification of radioactive liquid wastes, Treatment options for spent resins and concentrates - 16405

    International Nuclear Information System (INIS)

    Roth, Andreas

    2009-01-01

    Ion exchange is one of the most common and effective treatment methods for liquid radioactive waste. However, spent ion exchange resins are considered to be problematic waste that in many cases require special approaches and pre-conditioning during its immobilization to meet the acceptance criteria for disposal. Because of the function that they fulfill, spent ion exchange resins often contain high concentrations of radioactivity and pose special handling and treatment problems. Another very common method of liquid radioactive waste treatment and water cleaning is the evaporation or diaphragm filtration. Both treatment options offer a high volume reduction of the total volume of liquids treated but generate concentrates which contain high concentrations of radioactivity. Both mentioned waste streams, spent resins as well as concentrates, resulting from first step liquid radioactive waste treatment systems have to be conditioned in a suitable manner to achieve stable waste products for final disposal. Spent resin and concentrate treatment often appear as a specific task in decommissioning projects, because in the past those waste streams typically had been stored in tanks for the lifetime of the plant and needs to be retrieved, conditioned and packed prior to dismantling activities. Additionally a large amount of contaminated liquids will be generated by utilizing decontamination processes and needs to be processed further on. Such treatment options need to achieve waste products acceptable for final disposal, because due to the closure of the site no interim storage can be envisaged. The most common method of treatment of such waste streams is the solidification in a solid matrix with additional inactive material like cement, polymer etc. In the past good results have been achieved and the high concentration of radioactivity can be reduced by adding the inactive material. On the other hand, under the environment of limited space for interim storage and the absence

  5. Spent fuel and HLW transportation the French experience

    International Nuclear Information System (INIS)

    Giraud, J.P.; Charles, J.L.

    1995-01-01

    With 53 nuclear power plants in operation at EDF and a fuel cycle with recycling policy of the valuable materials, COGEMA is faced with the transport of a wide range of radioactive materials. In this framework, the transport activity is a key link in closing the fuel cycle. COGEMA has developed a comprehensive Transport Organization System dealing with all the sectors of the fuel cycle. The paper will describe the status of transportation of spent fuel and HLW in France and the experience gathered. The Transport Organization System clearly defines the role of all actors where COGEMA, acting as the general coordinator, specifies the tasks to be performed and brings technical and commercial support to its various subcontractors: TRANSNUCLEAIRE, specialized in casks engineering and transport operations, supplies packaging and performs transport operations, LEMARECHAL and CELESTIN operate transport by truck in the Vicinity of the nuclear sites while French Railways are in charge of spent fuel transport by train. HLW issued from the French nuclear program is stored for 30 years in an intermediate storage installation located at the La Hague reprocessing plant. Ultimately, these canisters will be transported to the disposal site. COGEMA has set up a comprehensive transport organization covering all operational aspects including adapted procedures, maintenance programs and personnel qualification

  6. Studies and research concerning BNFP. Nuclear spent fuel transportation studies

    International Nuclear Information System (INIS)

    Anderson, R.T.; Maier, J.B.

    1979-11-01

    Currently, there are a number of institutional problems associated with the shipment of spent fuel assemblies from commercial nuclear power plants: new and conflicting regulations, embargoing of certain routes, imposition of transport safeguards, physical security in-transit, and a lack of definition of when and where the fuel will be moved. This report presents a summary of these types and kinds of problems. It represents the results of evaluations performed relative to fuel receipt at the Barnwell Nuclear Fuel Plant. Case studies were made which address existing reactor sites with near-term spent fuel transportation needs. Shipment by either highway, rail, water, or intermodal water-rail was considered. The report identifies the impact of new regulations and uncertainty caused by indeterminate regulatory policy and lack of action on spent fuel acceptance and storage. This stagnant situation has made it impossible for industry to determine realistic transportation scenarios for business planning and financial risk analysis. A current lack of private investment in nuclear transportation equipment is expected to further prolong the problems associated with nuclear spent fuel and waste disposition. These problems are expected to intensify in the 1980's and in certain cases will make continuing reactor plant operation difficult or impossible

  7. Baseline descriptions for LWR spent fuel storage, handling, and transportation

    International Nuclear Information System (INIS)

    Moyer, J.W.; Sonnier, C.S.

    1978-04-01

    Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables

  8. Accurate determination of 129I concentrations and 129I/137Cs ratios in spent nuclear resins by Accelerator Mass Spectrometry

    International Nuclear Information System (INIS)

    Nottoli, Emmanuelle; Bienvenu, Philippe; Labet, Alexandre; Bourlès, Didier; Arnold, Maurice; Bertaux, Maité

    2014-01-01

    Determining long-lived radionuclide concentrations in radioactive waste has fundamental implications for the long-term management of storage sites. This paper focuses on the measurement of low 129 I contents in ion exchange resins used for primary fluid purification in Pressurised Water Reactors (PWR). Iodine-129 concentrations were successfully determined using Accelerator Mass Spectrometry (AMS) following a chemical procedure which included (1) acid digestion of resin samples in HNO 3 /HClO 4 , (2) radioactive decontamination by selective iodine extraction using a new chromatographic resin (CL Resin), and (3) AgI precipitation. Measured 129 I concentrations ranged from 4 to 12 ng/g, i.e. from 0.03 to 0.08 Bq/g. The calculation of 129 I/ 137 Cs activity ratios used for routine waste management produced values in agreement with the few available data for PWR resin samples. - Highlights: • In the context of radioactive waste management, this study aimed at measuring 129 I in spent resins using accelerator mass spectrometry. • The treatment procedure included microwave acid digestion of samples, iodine extraction by CL resins and AgI precipitation. • Developed first on synthetic matrices, the chemical treatment procedure was then successfully applied to real resin samples. • 129 I concentrations ranged from 4 to 12 ng/g of dry resin. • Results are in agreement with previous measurements and support reference values currently used for nuclear resin management

  9. Risk assessment in spent fuel storage and transportation

    International Nuclear Information System (INIS)

    Pandimani, S.

    1989-01-01

    Risk assessment in various stages of nuclear fuel cycle is still an active area of Nuclear safety studies. From the results of risk assessment available in literature, it can be determined that the risk resulting from shipments of plutonium and spent-fuel are much greater than that resulting from the transport of other materials within the nuclear fuel cycle. In India spent fuels are kept in Spent Fuel Storage Pool (SFSP) for about 240-400 days, which is relatively a longer period compared to the usual 120 days as recommended by regulatory authorities. After cooling spent fuels are transported to the reprocessing sites which are mostly situated close to the plants. India has two high level waste treatment facilities, one PREFRE (Plutonium Reprocessing and Fuel Recycling) at Tarapur and the other one, a unit of Nuclear Fuel Complex at Hyderabad. This paper presents the risk associated with spent fuel storage and transportation for the Indian conditions. All calculations are based on a typical CANDU reactor system. Simple fault tree models are evolved for SFSP and for Transportation Accident Mode (TAM) for both road and rail. Fault tree quantification and risk assessment are done to each of these models. All necessary data for SFSP are taken mostly from Reactor Safety Study, (1975). Similarly, the data for rail TAM are taken from Annual Statistical Statements, (1987-8) and that for road TAM from Special Issue on Motor Vehicle Accident Statistics in India, (1986). Simulation method is used wherever necessary. Risk is also estimated for normal/accident free transport

  10. Reclaimability of the spent sand mixture – sand with bentonite – sand with furfuryl resin

    Directory of Open Access Journals (Sweden)

    J. Dańko

    2011-04-01

    Full Text Available Introduction of new binding materials and new technologies of their hardening in casting moulds and cores production requires theapplication of reclamation methods adequate to their properties as well as special devices realizing tasks. The spent sands circulationsystem containing the same kind of moulding and core sands is optimal from the point of view of the expected reclamation results.However, in the face of a significant variability of applied technologies and related to them various reclamation methods, the need - of theobtained reclamation products assessment on the grounds of systematic criteria and uniform bases – arises, with a tendency of indicatingwhich criteria are the most important for the given sand system. The reclaimability results of the mixture of the spent moulding sand withGeko S bentonite and the spent core sand with the Kaltharz 404U resin hardened by acidic hardener 100 T3, are presented in the paper.Investigations were performed with regard to the estimation of an influence of core sands additions (10 –25% on the reclaimed materialquality. Dusts and clay content in the reclaim, its chemical reaction (pH and ignition loss were estimated. The verification of the reclaiminstrumental assessment was performed on the basis of the technological properties estimation of moulding sand with bentonite, where the reclaimed material was used as a matrix.

  11. Leaching behavior of 60Co and 137Cs from spent ion exchange resins in cement-bentonite clay matrix

    Science.gov (United States)

    Plecas, Ilija; Pavlovic, Radojko; Pavlovic, Snezana

    2004-05-01

    The leach rate of 60Co and 137Cs from two different ion exchange resins: (a) spent cation exchange resins and (b) spent mix bead ion exchange resins in cement-bentonite matrix has been studied. The solidification matrix was a standard Portland cement mixed with 290-350 kg/m 3 spent cation exchange resins, with or without 2-5% of bentonite clay. The leach rates from the cement-bentonite matrix as 60Co: (4.2-7.3) × 10 -5 cm/d, and for 137Cs: (3.2-6.6) × 10 -5 cm/d, after 245 days were measured. From the leaching data the apparent diffusivity of cobalt and cesium in cement-bentonite clay matrix with a waste load of 290-350 kg/m 3 spent cation exchange resins was measured as 60Co: (1.0-4.0) × 10 -6 cm 2/d and for 137Cs: (0.5-2.6) × 10 -4 cm 2/d after 245 days. These results are part of a 20-year mortar and concrete testing project which will influence the design of radioactive waste management for a future Serbian radioactive waste disposal center.

  12. Exorcising spent fuel transportation using comparative hazard assessment methods

    Energy Technology Data Exchange (ETDEWEB)

    Pennington, Charles W. [NAC international, Norcross (United States)

    2003-07-01

    Spent fuel transportation has achieved an exemplary safety record over more than three decades within both the United States (U.S.) and the global community at large. Today, many groups are attempting to precipitate fear of spent fuel transportation within the general public by 'demonizing' this proven technology and by creating a highly charged environment of radiation phobia. The actions of these groups within the U.S. result from the confluence of the terrorist acts of September 11, 2001, and the acceptance by the President and Congress of the U.S. Department of Energy's (DOE) recommendation of Yucca Mountain as the repository site for the disposal of the nation's spent fuel. This paper offers a comparative hazard assessment demonstrating the relative safety of spent fuel transportation in the context of currently accepted practices within society to show that there are no 'demons' associated with spent fuel transportation. The paper provides an assessment of potential population exposures based on more than 25 years of transport cask analysis and testing under beyond-design-basis (BDB) event conditions, including missile attacks, with those from current accepted activities within society that produce high dose exposures to the general public. Over the last quarter of a century, several spent fuel cask test programs have produced data that allow calculation of potential releases and population doses resulting from a terrorist attack. The DOE has used this information to develop projected worst-case population exposures as part of the Final Environmental Impact Statement (FEIS) for the Yucca Mountain repository. The paper discusses these potential releases and population exposures. Additionally, the paper identifies current unregulated activities and practices within societies yielding population exposures that exceed significantly those that would result from such highly hypothetical and improbable events as a terrorist missile

  13. Exorcising spent fuel transportation using comparative hazard assessment methods

    International Nuclear Information System (INIS)

    Pennington, Charles W.

    2003-01-01

    Spent fuel transportation has achieved an exemplary safety record over more than three decades within both the United States (U.S.) and the global community at large. Today, many groups are attempting to precipitate fear of spent fuel transportation within the general public by 'demonizing' this proven technology and by creating a highly charged environment of radiation phobia. The actions of these groups within the U.S. result from the confluence of the terrorist acts of September 11, 2001, and the acceptance by the President and Congress of the U.S. Department of Energy's (DOE) recommendation of Yucca Mountain as the repository site for the disposal of the nation's spent fuel. This paper offers a comparative hazard assessment demonstrating the relative safety of spent fuel transportation in the context of currently accepted practices within society to show that there are no 'demons' associated with spent fuel transportation. The paper provides an assessment of potential population exposures based on more than 25 years of transport cask analysis and testing under beyond-design-basis (BDB) event conditions, including missile attacks, with those from current accepted activities within society that produce high dose exposures to the general public. Over the last quarter of a century, several spent fuel cask test programs have produced data that allow calculation of potential releases and population doses resulting from a terrorist attack. The DOE has used this information to develop projected worst-case population exposures as part of the Final Environmental Impact Statement (FEIS) for the Yucca Mountain repository. The paper discusses these potential releases and population exposures. Additionally, the paper identifies current unregulated activities and practices within societies yielding population exposures that exceed significantly those that would result from such highly hypothetical and improbable events as a terrorist missile attack on a spent fuel

  14. Transportation of spent fuel from light water reactors

    International Nuclear Information System (INIS)

    Bernard, H.

    1993-01-01

    The French 'Compagnie Generale des Matieres Nucleaires' - COGEMA - is involved in the whole nuclear fuel cycle about 20 years. Among the different parts of the cycle, the Transport of Radioactive Materials, acting as a link between the differents plants has a great importance. As nuclear material transportation is the only fuel cycle step to be performed on public grounds, the industrial task has to be performed with the utmost stringent safety criteria. COGEMA and associates is now operating a fully mature commercial activity, with some 300 spent fuel shipments per year from its reprocessing customer's reactors to the LA HAGUE plant, either by rail, road or sea. The paper will review the organization of COGEMA transportation business, the level of technology with an update of the casks used for spent fuel, and the operational experience, with a particular view of the maintenance policy. (author)

  15. Oxidative degradation of low and intermediate level Radioactive organic wastes 2. Acid decomposition on spent Ion-Exchange resins

    International Nuclear Information System (INIS)

    Ghattas, N.K.; Eskander, S.B.

    1995-01-01

    The present work provides a simplified, effective and economic method for the chemical decomposition of radioactively contaminated solid organic waste, especially spent ion - exchange resins. The goal is to achieve volume reduction and to avoid technical problems encountered in processes used for similar purposes (incineration, pyrolysis). Factors efficiency and kinetics of the oxidation of the ion exchange resins in acid medium using hydrogen peroxide as oxidant, namely, duration of treatment and the acid to resin ratio were studied systematically on a laboratory scale. Moreover the percent composition of the off-gas evolved during the decomposition process was analysed. 3 figs., 5 tabs

  16. Safety criteria for spent-fuel transport. Final report

    International Nuclear Information System (INIS)

    Goldmann, K.; Gekler, W.C.

    1986-10-01

    The focus of this study is on the question, ''Do current regulations provide reasonable assurance of safety for a transport scenario of spent fuel, as presently anticipated by the Department of Energy, under the Nuclear Waste Policy Act.'' This question has been addressed by developing a methodology for identifying the expected frequency of Accidents Which Exceed Regulatory Conditions in Severity (AWERCS) for spent fuel transport casks and then assessing the health effects resulting from that frequency. By applying the methodology to an illustrative case of road transports, it was found that the accidental release of radioactive material from impact AWERCS would make negligible contributions to health effects associated with spent fuel transports by road. It is also concluded that the current regulatory drop test requirements in 10 CFR 71.51 which form the basis for cask design and were used to establish AWERCS screening criteria for this study are adequate, and that no basis was found to conclude that cask performance under expected road accident conditions represents an undue risk to the public

  17. Thermal analysis of transportation packaging for nuclear spent fuel

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki

    1989-01-01

    Safety analysis of transportation packaging for nuclear spent fuel comprises structural, thermal, containment, shielding and criticality factors, and the safety of a packaging is verified by these analyses. In thermal analysis, the temperature of each part of the packaging is calculated under normal and accident test conditions. As an example of thermal analysis, the temperature distribution of a packaging being subjected to a normal test was calculated by the TRUMP code and compared with measured data. (author)

  18. Present status and prospect of spent fuel transportation

    International Nuclear Information System (INIS)

    Adachi, H.

    1987-01-01

    Problems linked with spent fuel transportation in Japan, where there are 35 NPPs in operation, are considered. Every year about 500 t U are shipped to fuel reprocessing plants in Japan, as well as in France and UK. Four kinds of casks: HZ, EXCELLOX, TN and TK - are used for this purpose. By the mid-1990's it is suggested to build in Japan fuel reprocessing plant with capacity of 800 t U per year

  19. Simultaneous separation and determination of {sup 14}C and {sup 3}H in spent resins from PWR nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soon Dal; Kim, Jung Suck; Kim, Jong Goo; Han, Sun Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-09-15

    In this work {sup 14}C and {sup 3}H distribution characteristics of spent resins from nuclear power plants(NPPs), pressurized water reactors(PWRs), was investigated. It was found that the recovery percent of {sup 14}C by the wet oxidation-acid stripping was 81% {approx} 100% for the added activity range of {sup 14}C, 72 Bq {approx} 460 Bq, and it was not affected by the kinds of stripping acids, 3 N-HCI, 3 N-HNO{sub 3} and 3 N-H{sub 2}SO{sub 4}. And the recovery percent of {sup 3}H by distillation using the same apparatus was 81% {approx} 101% for the added activity range of {sup 3}H, 0.60 Bq {approx} 435 Bq. Among the tested stripping acids, 3 N-HCI, 3 N-NHO{sub 3} and 3 N-H{sub 2}SO{sub 4}, only the trapped {sup 3}H solution by distillation in 3 N-H{sub 2}SO{sub 4} was compatible with the 3H scintillator, Ultimagold XR. Neither of the {sup 14}C and {sup 3}H trapping solutions from the spent ion exchange resin samples by the wet oxidation-3 N-H{sub 2}SO{sub 4} stripping contained gamma nuclides. However, some gamma nuclides, {sup 60}Co, {sup 134}Cs, {sup 137}Cs and {sup 54}Mn, were found in the trapped {sup 3}H solutions of the spent resins by the wet oxidation-3 N-HCI stripping. It was the same for the {sup 3}H trapping solutions of the spent resins by Sample Oxidizer (PACKARD MODEL 307). Meanwhile only two nuclides, {sup 134}Cs, and {sup 137}Cs, were found in the {sup 14}C trapping solutions of the spent resins by Sample Oxidizer(PACKARD MODEL 307). It was found that most of the {sup 14}C in the spent resins existed as inorganic carbon form, more than about 70% of the total {sup 14}C center. Among the analyzed 30 spent ion exchange resin samples, the average concentration of {sup 14}C and {sup H} for the high radioactive samples, 8 samples, was 19000 Bq/g {+-} 41000 Bq/g, 670 Bq/g {+-} 460 Bq/g and that for the low radioactive samples, 22 samples, was 4.2 Bq/g {+-} 4.3 Bq/g, 6.0 Bq/g {+-} 5.3 Bq/g, respectively.

  20. Cementation of secondary wastes generated from carbonisation of spent organic ion exchange resins from nuclear power plants

    International Nuclear Information System (INIS)

    Sathi Sasidharan, N.; Deshingkar, D.S.; Wattal, P.K.

    2004-07-01

    The spent IX resins containing radioactive fission and activation products from power reactors are highly active solid wastes generated during operations of nuclear reactors. Process for carbonization of IX resins to achieve weight and volume reduction has been optimized on 50 dm 3 /batch pilot test rig. The process generates carbonaceous residue, organic liquid condensates (predominantly styrene) and aqueous alkaline scrubber solutions as secondary wastes. The report discusses laboratory tests on leaching of 137 Cs from cement matrix incorporating carbonaceous residues and extrapolation of results to 200 liter matrix block. The cumulative fraction of 137 Cs leached from 200 liter cement matrix was estimated to be 0.0021 in 200 days and 0.0418 over a period of 30 years. Incorporation of organic liquid condensates into cement matrix has been tried out successfully. Thus two types of secondary wastes generated during carbonization of spent IX resins can be immobilized in cement matrix. (author)

  1. Integrated spent fuel storage and transportation system using NUHOMS

    International Nuclear Information System (INIS)

    Lehnert, R.; McConaghy, W.; Rosa, J.

    1990-01-01

    As utilities with nuclear power plants face increasing near term spent fuel store needs, various systems for dry storage such as the NUTECH Horizontal Modular Storage (NUHOMS) system are being implemented to augment existing spent fuel pool storage capacities. These decisions are based on a number of generic and utility specific considerations including both short term and long term economics. Since the US Department of Energy (DOE) is tasked by the Nuclear Waste Policy Act with the future responsibility of transporting spent fuel from commercial nuclear power plants to a Monitored Retrievable Storage (MRS) facility anchor a permanent geologic repository, the interfaces between the utilities at-reactor dry storage system and the DOE's away-from-reactor transportation system become important. This paper presents a study of the interfaces between the current at-reactor NUHOMS system and the future away-from-reactor DOE transportation system being developed under the Office of Civilian Radioactive Waste Management (OCRWM) program. 7 refs., 9 figs., 1 tab

  2. Probabilistic safety analysis of transportation of spent fuel

    International Nuclear Information System (INIS)

    Subramaniam, Chitra

    1999-11-01

    The report presents the results of the study carried out to estimate the accident risk involved in the transport of spent fuel from Rajasthan Atomic Power Station near Kota to the fuel reprocessing plant at Tarapur. The technique of probabilistic safety analysis is used. The fuel considered is the Indian pressurised heavy water reactor fuel with a minimum cooling period of 485 days. The spent fuel is transported in a cuboidal, naturally-cooled shipping cask over a distance of 822 km by rail. The Indian rail accident statistics are used to estimate the basic rail accident frequency. The possible ways in which a release of radioactive material can occur from the spent fuel cask are identified by the fault tree analysis technique. The release sequences identified are classified into eight accident severity categories, and release fractions are assigned to each. The consequences resulting from the release are estimated by the computer code RADTRAN 4. Results of the risk analysis indicate that the accident risk values are very low and hence acceptable. Parametric studies show that the risk would continue to be small even if the controlling parameters were to simultaneously take extreme adverse values. (author)

  3. Thirty years of transport package development for spent fuels

    International Nuclear Information System (INIS)

    Cory, A.R.

    2005-01-01

    By June 2005, when shipments of spent fuel for reprocessing from Germany are concluded, BNFL flask types will have been responsible for transporting more than 2000 tonnes of heavy metal in Europe in the form of spent fuel. Several thousand more tonnes of spent fuel have been transported by sea from Japan over the last thirty years. The design of spent fuel packages has not stood still for that time. In order to anticipate the changing needs of the nuclear power generation industry, advances have been made both in package design and analysis. Thirty years ago spent fuel burnup and initial enrichment were considerably lower, which was reflected in the different demands placed on the shielding design of packages, and in the design of the internal basket to separate the fuel assemblies. Technical development of both 'wet' (water-filled cavity) and 'dry' packages has progressed in parallel, and the relative merits and peculiarities of each type is explored. BNFL has considerable experience in the operation of both types, and is well placed to comment on practical and functional issues associated with both types. While there have been certain evolutionary changes affecting package design, there have also been more significant changes in the Design Safety Case. These have sometimes been necessary to meet changes in IAEA Regulations, or the challenges posed by the regulators themselves. In other cases advantage has been taken of improvements in analytical techniques to demonstrate increased margins of operational safety. Where possible these margins have also been increased by other means, such as taking advantage of commercial trends to reduce package thermal loads. A key factor over the last thirty years has been the increasing influence of the Regulating Authorities and the development of the IAEA Regulations. The various Competent Authorities now tend to have a higher proportion of technical experts, often recruited from the nuclear industry, and are thus more able to

  4. Spent fuel transportation cask response to a tunnel fire scenario

    International Nuclear Information System (INIS)

    Bajwa, C.S.; Adkins, H.E.; Cuta, J.M.

    2004-01-01

    On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB), the U.S. agency responsible for determining the cause of transportation accidents, to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for Nuclear Waste Regulatory Analyses (CNWRA), and Pacific Northwest National Laboratory (PNNL) to determine the thermal conditions that existed in the Howard Street tunnel fire and analyze the effects of this fire on various spent fuel transportation cask designs. The Fire Dynamics Simulator (FDS) code, developed by NIST, was used to determine the thermal environment present in the Howard Street tunnel during the fire. The FDS results were used as boundary conditions in the ANSYS registered and COBRA-SFS computer codes to evaluate the thermal performance of different cask designs. The staff concluded that the transportation casks analyzed would withstand a fire with thermal conditions similar to those that existed in the Baltimore tunnel fire event. No release of radioactive materials would result from exposure of the casks analyzed to such an event. This paper describes the methods and approach used for this assessment

  5. Policy issues of transporting spent nuclear fuel by rail

    International Nuclear Information System (INIS)

    Spraggins, H.B.

    1994-01-01

    The topic of this paper is safe and economical transportation of spent nuclear fuel by rail. The cost of safe movement given the liability consequences in the event of a rail accident involving such material is the core issue. Underlying this issue is the ability to access the risk probability of such an accident. The paper delineates how the rail industry and certain governmental agencies perceive and assess such important operational, safety, and economic issues. It also covers benefits and drawbacks of dedicated and regular train movement of such materials

  6. Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Carlsen, Brett; Chiguer, Mustapha; Grahn, Per; Sampson, Michele; Wolff, Dietmar; Bevilaqua, Arturo; Wasinger, Karl; Saegusa, Toshiari; Seelev, Igor

    2016-09-01

    Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years until reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on

  7. Safety and Licensing of Spent Fuel Storage and Transport — Safety Issues Within Spent Fuel Transport

    International Nuclear Information System (INIS)

    Brut, S.; Derlot, F.; Milet, L.

    2015-01-01

    We can consider the different safety issues within French fuel transport as follows: (a) the proof as regards the leaking fuel assembly transport with hydrogen generation coming from potential in leakage water inside fuel rods; ( b) the measures taken to enforce the new design as well as the new manufacturing which have been decided since January 1 st 2007 in the frame of the 96 IAEA Regulation as regards the full water penetration as compared to the 85 IAEA Regulation, the latter allowing partial water penetration on certain conditions; and (c) the obligation of implementing various risk controls on exploitation site in order to take into account the possible human failure which are intrinsically increasing the permissible doses rates for workers. Even quite recently the leaking fuel assembly transport has been considered with no specific measure as regards the radiolysis phenomenon or the quality of drying cask holds. All these measures were sufficiently in accordance to rule out this issue. Lately, the leaking fuel assembly transport needs the implementation of equipment controls involved in nuclear power plants as regards the hydrogen rate before loading departure in order to determine on the evolution law, the maximum duration authorized for the transportation to not exceed the lower limit of inflammable status. As regards the proof of the criticality-safety casks, the main justification to be held on the irradiated fuel assembly on drop accident conditions could find a key in the hypothesis of the important damage of the fuel but should be in this matter, compensated by a limit of containment penetration for safety reason. For this case, the application of the 96 IAEA Regulation involves the use of independent leak tightness barriers. TN International is introducing different examples in France linked to the selection of multiple barriers. When limited in leakage quantity of water inside the cask is considered for the criticality studies, the French Competent

  8. Incineration of ion-exchange resins in fluidized bed. Part of a coordinated programme on treatment of spent ion exchange resins

    International Nuclear Information System (INIS)

    Valkianinen, M.

    1980-10-01

    Incineration of ion-exchange resins in a fluidized bed was studied on the pilot plant scale. The test programme performed consisted of the testing of various bed materials and finding the optimal conditions of incineration of spent resins. Granular resins were incinerated in an ethanol-water mixture. Incinernation converts the organic resin into inert oxide material, which can be solidified for instance with cement. The weight of the ash was 1...20% and the volume 2...30% of the original resins, which contained 15...25% moisture. When solidified with cement the volume of the ash-concrete is 4...22% of the concrete of equal compressive strength acquired by direct solidification. Water immersion and heat tests of solidified ash showed satisfactory results. The absorption of Cs and Co in various bed materials was studied by means of inactive tracer materials. Biotite and chamotte absorbed significantly, but this absorption does not drastically help on the off gas side. The sintering of the bed materials in the presence of sodium was studied. Corundum, chamotte and biotite have a safety limit of 5% sodium of the bed's weight at 850 0 C

  9. Testing of a transport cask for research reactor spent fuel

    International Nuclear Information System (INIS)

    Mourao, Rogerio P.; Silva, Luiz Leite da; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2011-01-01

    Since the beginning of the last decade three Latin American countries which operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half-scale model for MTR fuel constructed in Argentina and tested in Brazil. Two test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. Although the specimen has not successfully performed the tests, its overall performance was considered very satisfactory, and improvements are being introduced to the design. A third test sequence is planned for 2011. (author)

  10. Spent fuel and high-level radioactive waste transportation report

    Energy Technology Data Exchange (ETDEWEB)

    1989-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  11. Spent fuel and high-level radioactive waste transportation report

    Energy Technology Data Exchange (ETDEWEB)

    1990-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  12. Spent fuel and high-level radioactive waste transportation report

    International Nuclear Information System (INIS)

    1990-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ''comprehensive overview of the issues.'' This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list

  13. Spent fuel and high-level radioactive waste transportation report

    International Nuclear Information System (INIS)

    1989-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ''comprehensive overview of the issues.'' This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list

  14. Spent Fuel and High-Level Radioactive Waste Transportation Report

    International Nuclear Information System (INIS)

    1992-03-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ''comprehensive overview of the issues.'' This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list

  15. The resistance to impact of spent Magnox fuel transport flasks

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    This book completes the papers of the four-year programme of research and demonstrations embarked upon by the CEGB in 1981, culminating in the spectacular train crash at Old Dalby in July 1984. It explains the CEGB's operations in relation to the transportation of spent Magnox fuel. The public tests described in this book are more effective in improving public understanding and confidence than any amount of explanations could have been, raising the wider question of how best the scientific community can respond to the legitimate concerns of the man and woman in the street about the generating of electricity from nuclear power. The contents are: Taking care; irradiated fuel transport in the UK; programming for flask safety; the use of scale models in impact testing; flask analytical studies; drop test facilities; demonstration drop test; a study of flask transport impact hazards; impact of Magnox irradiated fuel transport flasks into rock and concrete; rail crash demonstration scenarios; horizontal impact testing of quarter scale flasks using masonry targets; horizontal crash testing and analysis of model flatrols; flatrol test; analysis of full scale impact into an abutment; analysis of primary impact forces in the train crash demonstration; horizontal impact tests of quarter scale Magnox flasks and stylised model locomotives; predictive estimates for behaviour in the train crash demonstration; design and organization of the crash; execution of the crash demonstration by British Rail; instrumentation for the train crash demonstration; photography for the crash demonstration; a summary of the CEGB's flask accident impact studies

  16. Data module development for spent nuclear fuel transportation risk assessment

    International Nuclear Information System (INIS)

    Chen, S.Y.; Yuan, Y.C.; Zielen, A.J.

    1992-01-01

    In 1986, in compliance with the Nuclear Waste Policy Act of 1982, the U.S. Department of Energy (DOE) issued environmental assessments (EAs) of the potential repository sites for spent nuclear fuel (SNF) and high-level radioactive wastes. A major concern expressed in the public comments on the EAs was the need for a route-specific risk analysis. One approach to this concern was presented in 1987 by DOE's Office of Civilian Radioactive Waste Management (OCRWM) at a workshop on models for OCRWM transportation risk analysis in Salt Lake City, Utah. The DOE decided that analytical capabilities for a route-specific risk analysis should be maintained at the state level, thus requiring state-specific accident and agricultural output data. The DOE also decided that such an analysis warranted development of a more reliable data base for a number of key input parameters for transportation risk assessment. Such a data base would be compiled in a format acceptable for input into the existing transportation risk code RADTRAN. Seven data modules have been developed on the basis of these considerations. A system module was also developed for integrating the data modules and serves as a preprocessor of the RADTRAN input data file. The relationship of the modules to the RADTRAN input is shown

  17. Japanese perspectives and research on packaging, transport and storage of spent fuel

    International Nuclear Information System (INIS)

    Saegusa, T.; Ito, C.; Yamakawa, H.; Shirai, K.

    2004-01-01

    The Japanese policy on spent fuel is reprocessing. Until, reprocessed, spent fuel shall be stored properly. This paper overviews current status of transport and storage of spent fuel with related research in Japan. The research was partly carried out under a contract of Ministry of Economy, Trade and Industry of the Japanese government

  18. Japanese perspectives and research on packaging, transport and storage of spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Saegusa, T.; Ito, C.; Yamakawa, H.; Shirai, K. [Central Research Inst. of Electric Power Industry (CRIEPI), Abiko (Japan)

    2004-07-01

    The Japanese policy on spent fuel is reprocessing. Until, reprocessed, spent fuel shall be stored properly. This paper overviews current status of transport and storage of spent fuel with related research in Japan. The research was partly carried out under a contract of Ministry of Economy, Trade and Industry of the Japanese government.

  19. Analysis of near-term spent fuel transportation hardware requirements and transportation costs

    International Nuclear Information System (INIS)

    Daling, P.M.; Engel, R.L.

    1983-01-01

    A computer model was developed to quantify the transportation hardware requirements and transportation costs associated with shipping spent fuel in the commercial nucler fuel cycle in the near future. Results from this study indicate that alternative spent fuel shipping systems (consolidated or disassembled fuel elements and new casks designed for older fuel) will significantly reduce the transportation hardware requirements and costs for shipping spent fuel in the commercial nuclear fuel cycle, if there is no significant change in their operating/handling characteristics. It was also found that a more modest cost reduction results from increasing the fraction of spent fuel shipped by truck from 25% to 50%. Larger transportation cost reductions could be realized with further increases in the truck shipping fraction. Using the given set of assumptions, it was found that the existing spent fuel cask fleet size is generally adequate to perform the needed transportation services until a fuel reprocessing plant (FRP) begins to receive fuel (assumed in 1987). Once the FRP opens, up to 7 additional truck systems and 16 additional rail systems are required at the reference truck shipping fraction of 25%. For the 50% truck shipping fraction, 17 additional truck systems and 9 additional rail systems are required. If consolidated fuel only is shipped (25% by truck), 5 additional rail casks are required and the current truck cask fleet is more than adequate until at least 1995. Changes in assumptions could affect the results. Transportation costs for a federal interim storage program could total about $25M if the FRP begins receiving fuel in 1987 or about $95M if the FRP is delayed until 1989. This is due to an increased utilization of federal interim storage facility from 350 MTU for the reference scenario to about 750 MTU if reprocessing is delayed by two years

  20. Analysis of near-term spent fuel transportation hardware requirements and transportation costs

    Energy Technology Data Exchange (ETDEWEB)

    Daling, P.M.; Engel, R.L.

    1983-01-01

    A computer model was developed to quantify the transportation hardware requirements and transportation costs associated with shipping spent fuel in the commercial nucler fuel cycle in the near future. Results from this study indicate that alternative spent fuel shipping systems (consolidated or disassembled fuel elements and new casks designed for older fuel) will significantly reduce the transportation hardware requirements and costs for shipping spent fuel in the commercial nuclear fuel cycle, if there is no significant change in their operating/handling characteristics. It was also found that a more modest cost reduction results from increasing the fraction of spent fuel shipped by truck from 25% to 50%. Larger transportation cost reductions could be realized with further increases in the truck shipping fraction. Using the given set of assumptions, it was found that the existing spent fuel cask fleet size is generally adequate to perform the needed transportation services until a fuel reprocessing plant (FRP) begins to receive fuel (assumed in 1987). Once the FRP opens, up to 7 additional truck systems and 16 additional rail systems are required at the reference truck shipping fraction of 25%. For the 50% truck shipping fraction, 17 additional truck systems and 9 additional rail systems are required. If consolidated fuel only is shipped (25% by truck), 5 additional rail casks are required and the current truck cask fleet is more than adequate until at least 1995. Changes in assumptions could affect the results. Transportation costs for a federal interim storage program could total about $25M if the FRP begins receiving fuel in 1987 or about $95M if the FRP is delayed until 1989. This is due to an increased utilization of federal interim storage facility from 350 MTU for the reference scenario to about 750 MTU if reprocessing is delayed by two years.

  1. Estimated consequences from severe spent nuclear fuel transportation accidents

    International Nuclear Information System (INIS)

    Arnish, J.J.; Monette, F.; LePoire, D.; Biwer, B.M.

    1996-01-01

    The RISKIND software package is used to estimate radiological consequences of severe accident scenarios involving the transportation of spent nuclear fuel. Radiological risks are estimated for both a collective population and a maximally exposed individual based on representative truck and rail cask designs described in the U.S. Nuclear Regulatory Commission (NRC) modal study. The estimate of collective population risk considers all possible environmental pathways, including acute and long-term exposures, and is presented in terms of the 50-y committed effective dose equivalent. Radiological risks to a maximally exposed individual from acute exposure are estimated and presented in terms of the first year and 50-y committed effective dose equivalent. Consequences are estimated for accidents occurring in rural and urban population areas. The modeled pathways include inhalation during initial passing of the radioactive cloud, external exposure from a reduction of the cask shielding, long-term external exposure. from ground deposition, and ingestion from contaminated food (rural only). The major pathways and contributing radionuclides are identified, and the effects of possible mitigative actions are discussed. The cask accident responses and the radionuclide release fractions are modeled as described in the NRC modal study. Estimates of severe accident probabilities are presented for both truck and rail modes of transport. The assumptions made in this study tend to be conservative; however, a set of multiplicative factors are identified that can be applied to estimate more realistic conditions

  2. Leaching of Co and Cs from spent ion exchange resins in cement ...

    Indian Academy of Sciences (India)

    Unknown

    2003-08-22

    Aug 22, 2003 ... process produces deionized water, as the radioactive ions are removed together with non radioactive ones. Ion ex- changers are resins that are polymers with cross-linking. (connections between long carbon chains in a polymer). The resin has active groups in the form of electrically charged sites. At these ...

  3. Transporting spent fuel and reactor waste in Sweden experience from 5 years of operation

    International Nuclear Information System (INIS)

    Dybeck, P.; Gustafsson, B.

    1990-01-01

    This paper reports that since the Final Repository for Reactor Waste, SFR, was taken into operation in 1988, the SKB sea transportation system is operating at full capacity by transporting spent fuel and now also reactor waste from the 12 Swedish reactors to CLAB and SFR. Transports from the National Research Center, Studsvik to the repository has recently also been integrated in the system. CLAB, the central intermediate storage for spent fuel, has been in operation since 1985. The SKB Sea Transportation System consists today of the purpose built ship M/s Sigyn, 10 transport casks for spent fuel, 2 casks for spent core components, 27 IP-2 shielded steel containers for reactor waste and 5 terminal vehicles. During an average year about 250 tonnes of spent fuel and 3 -- 4000 m 3 of reactor waste are transported to CLAB and SFR respectively, corresponding to around 30 sea voyages

  4. Method for digesting spent ion exchange resins and recovering actinides therefrom using microwave radiation

    Science.gov (United States)

    Maxwell, III, Sherrod L.; Nichols, Sheldon T.

    1999-01-01

    The present invention relates to methods for digesting diphosphonic acid substituted cation exchange resins that have become loaded with actinides, rare earth metals, or heavy metals, in a way that allows for downstream chromatographic analysis of the adsorbed species without damage to or inadequate elution from the downstream chromatographic resins. The methods of the present invention involve contacting the loaded diphosphonic acid resin with concentrated oxidizing acid in a closed vessel, and irradiating this mixture with microwave radiation. This efficiently increases the temperature of the mixture to a level suitable for digestion of the resin without the use of dehydrating acids that can damage downstream analytical resins. In order to ensure more complete digestion, the irradiated mixture can be mixed with hydrogen peroxide or other oxidant, and reirradiated with microwave radiation.

  5. Transportation and storage of foreign spent power reactor fuel

    International Nuclear Information System (INIS)

    1979-01-01

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

  6. Spent nuclear fuel transportation: public issues and answers

    International Nuclear Information System (INIS)

    Hoffman, W.D.

    1986-01-01

    The court-ordered shipping of 750 spent nuclear fuel assemblies from West Valley, New York back to their utility owners has generated considerable public and media interest. This paper discusses the specific concerns of the general public over the West Valley shipments, the issues raised by opposition groups, the interest of public officials and emergency preparedness teams as well as the media coverage generated. An analysis is performed on the effectiveness of the West Valley and utility public information programs utilized in addressing these issues, concerns and interests. Emphasis is placed on communications which work to facilitate the shipments and generate fuel transport acceptance. Information programs are discussed which increase preparedness for nuclear shipments by emergency response teams and build public confidence in their safety. The paper also examines communications which could have further enhanced the shipping campaign to date. Finally, plans are discussed for media preparation with interview training and press conferences. Emphasis is placed on materials provided for the media which have served to generate more favorable print and air time

  7. Bituminization of simulated waste, spent resins, evaporator concentrates and animal ashes by extrusion process

    International Nuclear Information System (INIS)

    Grosche Filho, C.E.; Chandra, U.

    1987-01-01

    The results of the study of bituminization of simulated radwaste - spennt ion-exchange resins, borate evaporator/concentrates and animal ashes, are presented and discussed. Distilled and oxidizer bitumen were used. Characterization of the crude material and simulated wastes-bitumen mixtures of varying weigt composition (30, 40, 50, 60% by weight of dry waste material) was carried out. The asphaltene and parafin contents in the bitumens were also determined. Some additives and were used with an aim to improve the characteristcs of solidified wastes. For leaching studies, granular ion-exchange resins were with Cs - 134 and mixtures of resin-bitumen were prepared. The leaching studies were executed using the IAEA recommendation and the ISO method. A conventional screw-extruder, common in plastic industry, was used determine operational parameters and process difficulties. Mixtures of resin-bitumen and evaporator concentrate-bitumen obtained from differents operational conditions were characterized. (Author) [pt

  8. Intermediate review on the transportation of spent fuel assemblies; Zwischenbilanz ueber die Transporte abgebrannter Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-15

    The transportation of spent fuel from the Swiss nuclear power plants to the reprocessing facilities in France and England was interrupted in May 1998 because of contamination that occurred. These measures were presented in the March 1999 statement made by the Office for the Safety of Nuclear Plants (HSK). The transport of spent fuel has been once more permitted and carried out under new conditions since August 1999. In its interim report of October 2000, HSK analyses and evaluates the experience gained since the resumption of transports. For each measure required, it compares the advantages and drawbacks and makes decisions on the maintenance or reduction of the measures to be taken. Between August 1999 and July 2000, 12 spent fuel transports were carried out between the Swiss nuclear power plants and the COGEMA reprocessing facility in France (7 from Goesgen, 4 from Beznau and 1 from Leibstadt). Neither noticeable disagreement with nor exceeding of contamination limits were noted during those 12 transports. This satisfactory result demonstrates that the measures required to be taken are effective. HSK expected from the measures a reduction of the frequency of exceeding contamination limits to less than 5% and also a marked reduction in their frequency. The present results correspond to this expectation; however, the statistical basis is not yet sufficient to be able to draw definitive conclusions. Nevertheless it is noticed that the situation in France, where similar measures have been taken, was very clearly improved. The frequency of exceeding contamination limits was reduced to 2% during the first semester of the year 2000, while it amounted to more than 30% before April 1998. It is the comprehensiveness of the measures required by HSK which allows the avoidance of contamination. The analysis shows that just a small number of measures only contribute insignificantly to the goal sought after. Therefore, two measures will be suppressed (packing of the empty

  9. Preparation of High Surface Area Activated Carbon from Spent Phenolic Resin by Microwave Heating and KOH Activation

    Science.gov (United States)

    Cheng, Song; Zhang, Libo; Zhang, Shengzhou; Xia, Hongying; Peng, Jinhui

    2018-01-01

    The spent phenolic resin is as raw material for preparing high surface area activated carbon (HSAAC) by microwave-assisted KOH activation. The effects of microwave power, activation duration and impregnation ratio (IR) on the iodine adsorption capability and yield of HSAAC were investigated. The surface characteristics of HSAAC were characterized by nitrogen adsorption isotherms, FTIR, SEM and TEM. The operating variables were optimized utilizing the response surface methodology (RSM) and were identified to be microwave power of 700 W, activation duration of 15 min and IR of 4, corresponding to a yield of 51.25 % and an iodine number of 2,384 mg/g. The pore structure parameters of the HSAAC, i. e., Brunauer-Emmett-Teller (BET) surface area, total pore volume, and average pore diameter were estimated to be 4,269 m2/g, 2.396 ml/g and 2.25 nm, respectively, under optimum conditions. The findings strongly support the feasibility of microwave-assisted KOH activation for preparation of HSAAC from spent phenolic resin.

  10. Preparation of titanates and zeolites and their uses in radioactive waste management, particularly in the treatment of spent resins

    International Nuclear Information System (INIS)

    Hultgren, Aa.; Airola, C.; Forsberg, S.; Faelt, L.

    1983-05-01

    Work on the preparation of titanates and zeolites was started and their possible uses in the management of radioactive wastes proposed in the first years of the 1970's by the Department of Nuclear Chemistry at the Royal Institute of Technology in Stockholm and the Chemical Center of the University of Lund, respectively. The major part of these efforts was sponsored by the National Council for Radioactive Waste (Prav), while the concluding phase including an over-all system design study adapted to the Swedish nuclear power programme for an economic evaluation was sponsored by the KBS Division of the Swedish Nuclear Fuel Supply Co. The preparation work resulted in processes ready for industrial scale production of sorbents of qualities adequate for applications in radioactive waste treatment. The essential effort was devoted to the treatment of spent resins from nuclear power plants by transfer to their radioactive contents to titanates and zeolites, which can then be sintered to stabel ceramic bodies (the PILO process). The economic evaluation indicated a significant economic incentive for the introduction of the PILO process if an incineration step is included for all types of spent resins. The essential efforts and results from this programme are summarized in this report. (Authors)

  11. Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix B, foreign research reactor spent nuclear fuel characteristics and transportation casks. Volume 2

    International Nuclear Information System (INIS)

    1995-03-01

    This is Appendix B of a draft Environmental Impact Statement (EIS) on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. It discusses relevant characterization and other information of foreign research reactor spent nuclear fuel that could be managed under the proposed action. It also discusses regulations for the transport of radioactive materials and the design of spent fuel casks

  12. Thermal analysis on NAC-STC spent fuel transport cask under different transport conditions

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Yumei [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Yang, Jian, E-mail: zdhjkz@zju.edu.cn [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Xu, Chao; Wang, Weiping [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Ma, Zhijun [Department of Material Engineering, South China University of Technology, Guangzhou (China)

    2013-12-15

    Highlights: • Spent fuel cask was investigated as a whole instead of fuel assembly alone. • The cask was successfully modeled and meshed after several simplifications. • Equivalence method was used to calculate the properties of parts. • Both the integral thermal field and peak values are captured to verify safety. • The temperature variations of key parts were also plotted. - Abstract: Transport casks used for conveying spent nuclear fuel are inseparably related to the safety of the whole reprocessing system for spent nuclear fuel. Thus they must be designed according to rigorous safety standards including thermal analysis. In this paper, for NAC-STC cask, a finite element model is established based on some proper simplifications on configurations and the heat transfer mechanisms. Considering the complex components and gaps, the equivalence method is presented to define their material properties. Then an equivalent convection coefficient is introduced to define boundary conditions. Finally, the temperature field is captured and analyzed under both normal and accident transport conditions by using ANSYS software. The validity of numerical calculation is given by comparing its results with theoretical calculation. Obtaining the integral distribution laws of temperature and peak temperature values of all vital components, the security of the cask can be evaluated and verified.

  13. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Lake, W.H.

    1989-01-01

    It is possible to develop an optimal strategy for implementing burnup credit in spent fuel transport casks. For transport, the relative risk is rapidly reduced if additional pre-transport controls such as a cavity dryness verifications are conducted prior to transport. Some other operational and design features that could be incorporated into a burnup credit cask strategy are listed. These examples represent many of the system features and alternatives already available for use in developing a broadly based criticality safety strategy for implementing burnup credit in the design and operation of spent fuel transport casks. 4 refs., 1 tab

  14. Comparison of extracted teeth and simulated resin blocks on apical canal transportation.

    Science.gov (United States)

    Khalilak, Zohreh; Fallahdoost, Arjang; Dadresanfar, Bahareh; Rezvani, Gholamreza

    2008-01-01

    We aimed to compare apical canal transportation of extracted teeth and two types of simulated resin blocks. Mesiobuccal root of extracted maxillary molars, high hardness simulated resin blocks (Knoop hardness=40) and low hardness simulated resin blocks (Knoop hardness=22) were prepared with K-files using step-back technique (n=15 canals in each group). Double exposure radiographic technique was used for extracted teeth. Simulated resin blocks were stabilized and scanned before and after preparation. Pre and post-preparation pictures were superimposed and apical transportation was measured. The data were analyzed statistically using ANOVA and Tukey HSD tests. There was no significant difference in apical canal transportation between extracted teeth and high hardness resin blocks (P>0.05). Low hardness resin blocks showed more apical transportation than the other groups (Presin blocks were similar.

  15. Experience of IEA-R1 research reactor spent fuel transportation back to United States

    Energy Technology Data Exchange (ETDEWEB)

    Frajndlich, Roberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Operacao do Reator IEAR-R1m]. E-mail: frajndli@net.ipen.br; Perrotta, Jose A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div.de Engenharia do Nucleo]. E-mail: perrotta@net.ipen.br; Maiorino, Jose Rubens [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Diretoria de Reatores]. E-mail: maiorino@net.ipen.br; Soares, Adalberto Jose [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Dept. de Reatores]. E-mail: ajsoares@net.ipen.br

    1998-07-01

    IPEN/CNEN-SP is sending the IEA-R1 Research Reactor spent fuels from USA origin back to this country. This paper describes the experience in organizing the negotiations, documents and activities to perform the transport. Subjects as cask licensing, transport licensing and fuel failure criteria for transportation are presented. (author)

  16. Improved management of SG BD demineralizer for reduced generation of low-level radioactive spent resin in Korean nuclear power plants

    International Nuclear Information System (INIS)

    Rhee, I.; Cho, D.; Yeon, J.

    2003-01-01

    Most nuclear power plants in Korea have adopted Ethanolamine(ETA) as a secondary pH control agent to increase the pH at the liquid phase, which may reduce the corrosion in steam generator tubes and moisture separator/reheat system. Along with its beneficial effect of SG protection from corrosion and degradation, the replacement of ammonia with ETA causes the increased generation of spent resin and the reduced run time of demineralizer in steam generator blowdown(SG BD) system. The composition ratio of cation- to anion- exchange resin in SG BD mixed bed should be increased in the ETA chemistry environment to meet the ratio of cation to anion in the aqueous solution, which results in the simultaneous exhaustion of cation and anion exchange resins. The utilization rate of mixed bed is greatest at the cation-to-anion ratio of 95:1 on the theoretical equivalent basis in the solution, but practically highest at that of 22:1 due to the possible inhomogeneous distribution of cation and anion exchange resins in SG BD bed. The run time of the bed could be extended by 30% such that, at that much, the purchase cost of new resin is saved and the production rate of spent resin is reduced. The guideline on the replacement of resin in SG BD bed is not necessary to secure the removal of radioactive particles without the leakage of the primary coolant into the secondary side since all the radioactive ions can be eliminated by SG BD bed with the sufficient time. They are retained during more than one month after their ingress into the SG BD bed without leakage. With the reduced replacement, thus, the SG BD spent resin that comprises 65% of low-level radioactive solid waste can be much cut down

  17. Review of the transportation of spent fuel assemblies; Bilanz ueber die Transporte abgebrannter Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-11-15

    In August 1999 the transportation of spent fuel assemblies from the Swiss nuclear power plants to the reprocessing facilities was resumed. Until October 2002, 37 transports were carried out without exceeding the legal limits on contamination and dose rate. 19 transports went to La Hague (4 from Beznau, 9 from Goesgen, 4 from Leibstadt and 2 from Muehleberg), 11 to BNFL in Sellafield (2 from Beznau and 9 from Muehleberg) and 7 to the Central Intermediate Storage for Radioactive Wastes (ZWILAG) in Wuerenlingen (4 from Goesgen and 3 from Leibstadt). Additionally, three deliveries with vitrified highly radioactive wastes from the reprocessing facilities were carried out to ZWILAG, also without exceeding any limits. These pleasing results show that the measures required to be taken, as required by the Federal Agency for the Safety of Nuclear Installations (HSK), have contributed to the safe and contamination-free accomplishment of such transportation. Even if the care taken in the carrying out of such transports may not lessen, HSK sees no reason to reduce the present technical and organisational measures. As far as the registration of the individual doses of the participating railway staff is concerned, the results obtained up till now confirm that the radiological exposure is negligible. From the point of view of radiological protection it is not necessary to pursue the registration of the individual radiological exposition. This is why the HSK no longer requires this measure any more. In France, too, the situation concerning contamination during transportation of spent fuel assemblies has been strikingly and lastingly improved. Contamination is only seldom observed and only in connection with a few identified nuclear power plants. In transfers between France and Germany, Belgium as well as the Netherlands, no exceeding of the contamination limits has been noted since resumption of transportation. During the 33 deliveries of empty containers and containers with

  18. Analyses of the transportation of spent research reactor fuel in the United States

    International Nuclear Information System (INIS)

    Cashwell, J.W.; Neuhauser, K.S.

    1989-01-01

    The Transportation Technology Center at Sandia National Laboratories has analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy (DOE) Office of Defense Programs. This effort represents the first comprehensive analytical evaluation of the risks of transporting high-, medium-, and low-enriched uranium spent research reactor fuel by both sea and land. Two separate shipment programs have been analyzed: the shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). In order to perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

  19. Development of a new neutron shielding material, TN trademark Resin Vyal for transport/storage casks for radioactive materials

    International Nuclear Information System (INIS)

    Abadie, P.

    2004-01-01

    TN trademark Resin Vyal, a patent pending composite, is a new neutron shielding material developed by COGEMA LOGISTICS for transport/storage casks of radioactive materials (spent fuel, reprocessed fuel,..). This material is composed of a thermosetting resin (vinylester resin in solution of styrene) and two mineral fillers (alumine hydrate and zinc borate). Its shielding ability for neutron radiation is related to a high hydrogen content (for slowing down neutrons) and a high boron content (for absorbing neutrons). Source of hydrogen is organic matrix (resin) and alumine hydrate; source of boron is zinc borate. Atomic concentrations are equal to 5.10 22 at/cm 3 for hydrogen and 9.10 20 at/cm 3 for boron. Due to the flame retardant character of components, the final material has a good fire resistance (it is auto-extinguishable). Its density is equal to 1,8. The manufacturing process of these material is easy: it consists in mixing the fillers and pouring in-situ (in cask); so, the curing of this composite, which leads to a three-dimensional structure, takes place at ambient temperature. Temperature resistance of this material was evaluated by performing exposition tests of samples at different temperatures (150 C to 170 C). According to tests results, its maximal temperature of use was confirmed equal to 160 C

  20. Analyses of the transportation of spent research reactor fuel in the United States

    International Nuclear Information System (INIS)

    Cashwell, J.W.; Neuhauser, K.S.

    1989-01-01

    We analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy's (DOE) Office of Defense Programs. Two separate shipment programs were analyzed. The shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). To perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. The risks of transporting spent nuclear fuel and other radioactive materials by all modes have been analyzed extensively. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

  1. Operation and maintenance of spent fuel storage and transportation casks/containers

    International Nuclear Information System (INIS)

    2007-01-01

    Member States have a growing need for casks for spent fuel storage and transportation. A variety of casks has been developed and is in use at an increasing number of sites. This has resulted in an accumulation of experience that will provide valuable information for other projects in spent fuel management. This publication provides a comprehensive review of information on the cask operation and maintenance associated with spent fuel storage. It draws upon generic knowledge from industrial experience and applications and is intended to serve as a basis for better planning and implementation in future projects

  2. U.S. spent fuel transportation security in the post 9/11 world

    International Nuclear Information System (INIS)

    Anne, Catherine; Patterson, John; Williams, Blake

    2002-01-01

    On September 11, 2002 the terrible tragedies in New York, Pennsylvania and Washington, DC changed the world forever. Security issues not only impact our daily lives, but are also in a state flux concerning the shipment of spent nuclear fuel in the United States. The formation of the Homeland Security Advisory System and Interim Compensatory Measures from the NRC, along with other security measures, have affected the way we transport spent nuclear fuel. This paper describes the challenging and demanding way that security is planned, implemented and maintained in support of spent fuel shipments in the United States. (author)

  3. Classification of transportation packaging and dry spent fuel storage system components according to importance to safety

    International Nuclear Information System (INIS)

    Tyacke, M.J.; McConnell, J.W. Jr.; Ayers, A.L. Jr.; O'Connor, S.C.; Jankovich, J.P.

    1996-01-01

    The Idaho National Engineering Laboratory prepared a technical report for the Office of Nuclear Material Safety and Safeguards of the US Nuclear Regulatory Commission, entitled Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety, NUREG/CR-6407. This paper provides the results of that report. It also presents the graded approach for classification of components used in transportation packagings and dry spent fuel storage systems. This approach provides a method for identifying the classification of components according to importance to safety within transportation packagings and dry spent fuel storage systems. Record retention requirements are discussed to identify the documentation necessary to validate that the individual components were fabricated in accordance with their assigned classification. A review of the existing regulations pertaining to transportation packagings and dry storage systems was performed to identify current requirements. The general types of transportation packagings and dry storage systems are identified. The methodology used in this paper is based on Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material. This paper also includes a list of generic components for each of the general types of transportation packagings and spent fuel storage systems, with a classification category assigned to each component. Several examples concerning the safety importance of components are presented

  4. Spent fuel shipping costs for transportation logistics analyses

    International Nuclear Information System (INIS)

    Cole, B.M.; Cross, R.E.; Cashwell, J.W.

    1983-05-01

    Logistics analyses supplied to the nuclear waste management programs of the U.S. Department of Energy through the Transportation Technology Center (TTC) at Sandia National Laboratories are used to predict nuclear waste material logistics, transportation packaging demands, shipping and receiving rates and transportation-related costs for alternative strategies. This study is an in-depth analysis of the problems and contingencies associated with the costs of shipping irradiated reactor fuel. These costs are extremely variable however, and have changed frequently (sometimes monthly) during the past few years due to changes in capital, fuel, and labor costs. All costs and charges reported in this study are based on January 1982 data using existing transport cask systems and should be used as relative indices only. Actual shipping costs would be negotiable for each origin-destination combination

  5. Below Grade Assessment of Spent Nuclear Fuel Cask Transport Route

    International Nuclear Information System (INIS)

    CHENAULT, D.M.

    1999-01-01

    The report provides an assessment of the route for the SNF Fuel transport system from the K Basins to the CVDF and to the CSB. Results include the identification of any underground structures or utilities traveled over by the transport, the overburden depths for all locations identified, evaluation of the loading conditions, and determination of the effects of the loads on the structures and utilities

  6. Development of a transport cask for spent fuel elements of research reactors

    International Nuclear Information System (INIS)

    Quintana, F.; Saliba, R.O.; Furnari, J.C.; Mourao, R.P; Leite da Silva, L.; Novara, O.; Alexandre Miranda, C.; Mattar Neto, M.

    2012-01-01

    This article presents an overview of the development of a research reactor spent fuel transport cask. Through a project funded by the IAEA, Argentina, Brazil and Chile have collaborated to enhance regional capacity in the management of spent fuel elements from research reactors operated in the region. A packaging for the transport of research reactors spent fuel was developed. It was designed by a team of researchers from the countries mentioned and a 1:2 scale model for MTR type fuel was constructed in Argentina and subsequently tested in CDTN facilities in Belo Horizonte, Brazil. There were three test sequences to test the cask for normal transport and hypothetical accident conditions. It has successfully passed the tests and the overall performance was considered satisfactory. As part of the licensing process, a test sequence with the presence of regulatory authorities is scheduled for December, 2012 (author)

  7. Transportation of failed or damaged foreign research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    Messick, C.E.; Mustin, T.P.; Massey, C.D.

    1998-01-01

    Since resuming the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy (DOE) and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues associated with the transport of Materials Testing Reactor (MTR)-type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be important to foreign research reactor operators, shippers, and cask vendors, so that appropriate amendments to the Certificate of Compliance for spent fuel casks can be submitted in a timely manner to facilitate the safe and scheduled transport of FRR SNF

  8. Safety of handling, storing and transportation of spent nuclear fuel and vitrified high-level wastes

    International Nuclear Information System (INIS)

    Ericsson, A.M.

    1977-11-01

    The safety of handling and transportation of spent fuel and vitrified high-level waste has been studied. Only the operations which are performed in Sweden are included. That is: - Transportation of spent fuel from the reactors to an independant spent fuel storage installation (ISFSI). - Temporary storage of spent fuel in the ISFSI. - Transportation of the spent fuel from the ISFSI to a foreign reprocessing plant. - Transportation of vitrified high-level waste to an interim storage facility. - Interim storage of vitrified high-level waste. - Handling of the vitrified high-level waste in a repository for ultimate disposal. For each stage in the handling sequence above the following items are given: - A brief technical description. - A description of precautionary measures considered in the design. - An analysis of the discharges of radioactive materials to the environment in normal operation. - An analysis of the discharges of radioactive materials due to postulated accidents. The dose to the public has been roughly and conservatively estimated for both normal and accident conditions. The expected rate of occurence are given for the accidents. The results show that above described handling sequence gives only a minor risk contribution to the public

  9. Treatment and conditioning of spent ion exchange resins from research reactors, precipitation sludges and other radioactive concentrates

    International Nuclear Information System (INIS)

    1993-02-01

    Spent radioactive ion exchange resins, precipitation sludges and evaporator concentrates are generated from the treatment of aqueous waste streams at nuclear research centres and from the use of radioisotopes in research and medical or industrial applications. A strategy for the effective management of these wastes from generation to disposal is necessary to ensure their safe handling, conditioning, storage and disposal to avoid detrimental effects on health and the environment. This report describes the factors to be considered in the development and selection of appropriate strategies for managing these types of wastes. Waste characteristics, pretreatment requirements, conditioning processes, packaging, and properties of the conditioned products are discussed. In addition, safety considerations and quality assurance needs are addressed. The report is intended to provide guidance to Member States that do not have nuclear power generation programmes. Several processes and procedures are presented, though preference is given to the simpler, easy-to-operate processes requiring relatively unsophisticated and inexpensive equipment. 24 refs, 20 figs, 13 tabs

  10. Assessment of health risks brought about by transportation of spent fuel; Kaeytetyn ydinpolttoaineen kuljetusten terveysriskien arviointi

    Energy Technology Data Exchange (ETDEWEB)

    Suolanen, V.; Lautkaski, R.; Rossi, J. [VTT Energy, Espoo (Finland)

    1999-03-01

    In the study health risks caused by transportation of spent fuel from Olkiluoto and from Loviisa NPP`s to the planned disposal site have been evaluated. The Olkiluoto NPP is owned by Teollisuuden Voima Oy (TVO) and the Loviisa NPP, situated at Haestholmen, by Fortum Power and Heat Oy. According to the base scenario of 40 years use of the current NPP`s the total amount of spent fuel will be 1840 tU (TVO) and 860 tU (Fortum). Annually, 110 tU on the average and at most 250 tU will be transported to the disposal site. The considered transportation routes are from Olkiluoto to Haestholmen, from Olkiluoto to Kivetty, from Olkiluoto to Romuvaara, from Haestholmen to Olkiluoto, from Haestholmen to Kivetty and from Haestholmen to Romuvaara. The considered transportation modes are truck, rail or ship, or combinations of these modes. Each transportation route has been divided into homogenised sequences with respect to population density and/or route type. Total amount of analysed route options were 40, some route sequences are overlapping. Radiation exposures to the population along the routes have been calculated in normal, incident and accident situations during transportation. Occupational radiation doses to the personnel have been estimated for normal transportation only. The consequences of normal transportation have been evaluated based on RADTRAN-model, developed by the Sandia National Laboratories. As incidents, stopping of spent fuel transportation for an exceptionally long period of time, and in another case contamination of outer surface of spent fuel cask have been considered. Expected collective doses and health risks of transportation accidents connected to the routes have been calculated with RADTRAN-model. Single hypothetical transport accidents with pessimistic release assumptions have been further analysed in more detail with the ARANO-model, developed by VTT (Technical Research Centre of Finland). (orig.) 9 refs.

  11. 77 FR 28406 - Spent Fuel Transportation Risk Assessment

    Science.gov (United States)

    2012-05-14

    ... Radioactive Material by Air and Other Modes,'' which assessed the adequacy of those regulations to provide... under routine and accident transport conditions, and the risk was found to be acceptable. Since that... accident. This investigation shows that the risk from the radiation emitted from the cask is a small...

  12. Probabilistic risk assessment on maritime spent nuclear fuel transportation (Part II: Ship collision probability)

    International Nuclear Information System (INIS)

    Christian, Robby; Kang, Hyun Gook

    2017-01-01

    This paper proposes a methodology to assess and reduce risks of maritime spent nuclear fuel transportation with a probabilistic approach. Event trees detailing the progression of collisions leading to transport casks’ damage were constructed. Parallel and crossing collision probabilities were formulated based on the Poisson distribution. Automatic Identification System (AIS) data were processed with the Hough Transform algorithm to estimate possible intersections between the shipment route and the marine traffic. Monte Carlo simulations were done to compute collision probabilities and impact energies at each intersection. Possible safety improvement measures through a proper selection of operational transport parameters were investigated. These parameters include shipment routes, ship's cruise velocity, number of transport casks carried in a shipment, the casks’ stowage configuration and loading order on board the ship. A shipment case study is presented. Waters with high collision probabilities were identified. Effective range of cruising velocity to reduce collision risks were discovered. The number of casks in a shipment and their stowage method which gave low cask damage frequencies were obtained. The proposed methodology was successful in quantifying ship collision and cask damage frequency. It was effective in assisting decision making processes to minimize risks in maritime spent nuclear fuel transportation. - Highlights: • Proposes a probabilistic framework on the safety of spent nuclear fuel transportation by sea. • Developed a marine traffic simulation model using Generalized Hough Transform (GHT) algorithm. • A transportation case study on South Korean waters is presented. • Single-vessel risk reduction method is outlined by optimizing transport parameters.

  13. Spent nuclear fuel and high level radioactive waste transportation. White paper

    International Nuclear Information System (INIS)

    1985-06-01

    The High-Level Radioactive Waste Committee of the Western Interstate Energy Board has been involved in a year-long cooperative project with the US Department of Energy (DOE) to develop an information base on the transportation of spent nuclear fuel and high-level radioactive waste (HLW) so that western states can be constructive and informed participants in the repository program under the Nuclear Waste Policy Act (NWPA). The historical safety record of transportation of HLW and spent fuel is excellent; no release of these radioactive materials has ever occurred during transportation. Projected shipments under the NWPA will, however, greatly exceed current shipments in the US. For example, over the past five years, 119 metric tons of civilian spent fuel have been shipped in this country, while shipments to the first and second repository are each expected to peak at 3000 metric tons per year. The Committee believes that the successful development and operation of a national HLW/spent fuel transportation system can best be accomplished through an open process based on the common sense approach of taking all reasonable measures to minimize public risk and performing whatever actions are reasonably required to promote public acceptance. Therefore, the Committee recommends that the Department of Energy further the goals of the NWPA by developing a Comprehensive Transportation Plan which adopts a systematic, comprehensive, and integrated approach to resolving all spent fuel and HLW transportation issues in a timely manner. The suggested scope of such a plan is discussed in this White paper. Many of the suggested elements of such a plan are similar to those being developed by the Department of energy for inclusion in the Department's Transportation Institutional Plan

  14. Assessment of the risk of transporting spent nuclear fuel by truck

    International Nuclear Information System (INIS)

    Elder, H.K.

    1978-11-01

    The assessment includes the risks from release of spent fuel materials and radioactive cask cavity cooling water due to transportation accidents. The contribution to the risk of package misclosure and degradation during normal transport was also considered. The results of the risk assessment have been related to a time in the mid-1980's, when it is projected that nuclear plants with an electrical generating capacity of 100 GW will be operating in the U.S. For shipments from reactors to interim storage facilities, it is estimated that a truck carrying spent fuel will be involved in an accident that would not be severe enough to result in a release of spent fuel material about once in 1.1 years. It was estimated that an accident that could result in a small release of radioactive material (primarily contaminated cooling water) would occur once in about 40 years. The frequency of an accident resulting in one or more latent cancer fatalities from release of radioactive materials during a truck shipment of spent fuel to interim storage was estimated to be once in 41,000 years. No accidents were found that would result in acute fatalities from releases of radioactive material. The risk for spent fuel shipments from reactors to reprocessing plants was found to be about 20% less than the risk for shipments to interim storage. Although the average shipment distance for the reprocessing case is larger, the risk is somewhat lower because the shipping routes, on average, are through less populated sections of the country. The total risk from transporting 180-day cooled spent fuel by truck in the reference year is 4.5 x 10 -5 fatalities. An individual in the population at risk would have one chance in 6 x 10 11 of suffering a latent cancer fatality from a release of radioactive material from a truck carrying spent fuel in the reference year

  15. Degradation of ion spent resin using the Fenton's reagent; Degradacao da resina de troca ionica utilizando o reagente de Fenton

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Leandro Goulart de

    2013-07-01

    The most common method for spent radioactive ion exchange resin treatment is its immobilization in cement, which reduces the radionuclides release into the environment. Although this method is efficient, it increases considerably the final volume of the waste due to the low incorporation capacity. The objective of this work was to develop a degradation method of spent resins arising from the nuclear research reactor located at the Nuclear and Energy Research Institute (IPEN-CNEN/SP), using an Advanced Oxidation Process (AOP) with Fenton's reagents. This method would allow a higher incorporation in cement. Three different resins were evaluated: cationic, anionic and a mixture of both resins. The reactions were conducted varying the catalyst concentration (25, 50, 100 and 150 mM), the volume of hydrogen peroxide (320 to 460 mL), and three different temperatures, 50, 60 and 70 deg C. Degradation of about 98% was achieved using a 50 mM catalyst solution and 330 mL of hydrogen peroxide solution. The most efficient temperature was 60 deg C. (author)

  16. Transportation of failed or damaged foreign research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    Messick, Charles E.; Mustin, Tracy P.; Massey, Charles D.

    1999-01-01

    Since initiating the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel causing a degradation of the fuel assembly exposing fuel meat and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues and implementation challenges associated with the transport of MTR type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, implementation status, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be of interest to foreign research reactor operators, shippers, and cask vendors in evaluating the condition of their fuel to ensure it can be transported in accordance with appropriate cask certificate requirements. (author)

  17. Transport experience of NH-25 spent fuel shipping cask for post irradiation examination

    International Nuclear Information System (INIS)

    Mori, Ryuji

    1982-01-01

    Since the Japan Atomic Energy Research Institute and Nippon Nuclear Fuel Development Co. hot laboratories are located far off from the port which can handle spent fuel shipping casks, it is necessary to use a trailer-mounted cask which can be transported by public roads, bridges and intersections for the transportation of spent fuel specimens to these hot laboratories. Model NH-25 shipping cask was designed, manufactured and oualification tested to meet Japanese regulations and was officially registered as a BM type cask. The NH-25 cask accomodates two BWR fuel assemblies, one PWR assembly or one ATR fuel assembly using interchangeable inner containers. The cask weight is 29.2 t. The cask has three concentric stainless steel shells. Gamma shielding is lead cast between the inner shell and the intermediate shell. Neutro n shielding consists of ethylene-glycol-aqueous solution layer formed between the intermediate shell and the outer shell. The NH-25 cask now has been in operation for 2.5 yr. It was used for the transportation of spent fuel assemblies from six LWR power plants to the port on shipping cask carrier ''Hinouramaru'' on the sea, as well as from the port to the hot laboratory on a trailer. The capability of safe handling and transporting of spent fuel assemblies has been well demonstrated. (author)

  18. Development of transport and storage cask for high burn-up spent fuel

    International Nuclear Information System (INIS)

    Kuri, S.; Tamaki, H.; Hode, S.

    2004-01-01

    Mitsubishi Heavy Industries, LTD. (MHI) has been developing various transport and storage casks (MSF cask fleet) for high burn-up spent nuclear fuel (SNF). This paper outlines the specifications and describes the features of the newly developed casks and the advanced technology that enables the maximize number of the accommodated fuel assemblies of high burn-up and short cooling period

  19. The operational and logistic experience on transportation of Brazilian spent fuel to USA

    International Nuclear Information System (INIS)

    Maiorino, Jose Rubens; Frajndlich, Roberto; Mandlae, Martin; Bensberg, Werner; Renger, August; Grabow, Karsten

    2000-01-01

    A shipment of 127 spent MTR fuel assemblies was made from IEA-R1 Research Reactor located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, Brazil to Savannah River Site Laboratory in the United States. This paper describes the operational and logistic experience on this transportation made by IPEN staff and the Consortium NCS/GNS. (author)

  20. Two-dimensional radiation shielding optimization analysis of spent fuel transport container

    International Nuclear Information System (INIS)

    Tian Yingnan; Chen Yixue; Yang Shouhai

    2013-01-01

    The intelligent radiation shielding optimization design software platform is a one-dimensional multi-target radiation shielding optimization program which is developed on the basis of the genetic algorithm program and one-dimensional discrete ordinate program-ANISN. This program was applied in the optimization design analysis of the spent fuel transport container radiation shielding. The multi-objective optimization calculation model of the spent fuel transport container radiation shielding was established, and the optimization calculation of the spent fuel transport container weight and radiation dose rate was carried by this program. The calculation results were checked by Monte-Carlo program-MCNP/4C. The results show that the weight of the optimized spent fuel transport container decreases to 81.1% of the origin and the radiation dose rate decreases to below 65.4% of the origin. The maximum deviation between the calculated values from the program and the MCNP is below 5%. The results show that the optimization design scheme is feasible and the calculation result is correct. (authors)

  1. Nickel plating of spent fuel element transportation and storage containers

    International Nuclear Information System (INIS)

    Bedenig, D.O.; Holly, F.

    1987-01-01

    For economic reasons, the stainless steel used for nuclear transport and storage containers has been replaced by a significantly less expensive, but almost equally suitable material - spheroidal graphite cast iron. Because of the impossibility of sufficiently decontaminating raw cast iron surfaces, a suitable coating had to be developed. Steel lining, epoxy painting and soft nickel plating are known possibilities. If both the requirements made on such a coating and the economics are considered, soft nickel seems to be the most attractive solution. The paper describes the process of soft nickel plating which was developed by Von Roll, Switzerland, based on its proprietary ''Toraxier-Process'' - in co-operation with GNS (Gesellschaft fuer Nuklear-Service mbH, Essen). Soft nickel plating has been successfully applied to more than 30 Castor containers. (author)

  2. Decontamination of transport casks and of spent fuel storage facilities

    International Nuclear Information System (INIS)

    1990-06-01

    The present document provides an analysis of the technical papers presented at the meeting as well as a summary of the panel discussion. Conclusions and Recommendations: The meeting agreed that the primary source of contamination of transport casks is the production of radioactive isotopes in nuclear fuel and activation products of fuel components in nuclear reactors. The type, amount of mechanism for the release of these isotopes depend on the reactor type and fuel handling process. The widespread use of pools for the storage and handling of fuel provides an easy path for the transfer of contamination. Control of pool water conditions is essential for limiting the spread of contamination. For plants where casks are immersed in pools for loading, the immersion times should be minimised. Casks should be designed for ease of decontamination. The meeting discussed the use of stainless steel and suitable paints for coating casks. Designers should consider the appropriate coating for specific applications. The use of pressurized water for decontamination is recommended whenever possible. A number of commercially available reagents exist for decontaminating cask external surfaces. More work, however, is needed to cope with Pressurized Water Reactor crud within casks. Leaking fuel should be identified and isolated before storage in pools. Basic studies of the uptake and release of contamination from cask surfaces should be initiated. Standardization of methods of contamination measurement and instrumentation should be instituted. Refs, figs and tabs

  3. HEU and Leu FueL Shielding Comparative Study Applied for Spent Fuel Transport

    International Nuclear Information System (INIS)

    Margeanu, C.A.; Margeanu, S.; Barbos, D.

    2009-01-01

    INR Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies. The dual core concept involves the operation of a 14 MW TRIGA steady-state, high flux research and material testing reactor at one end of a large pool, and the independent operation of an annular-core pulsing reactor (TRIGA-ACPR) at the other end of the pool. The steady-state reactor is mostly used for long term testing of power reactor fuel components (pellets, pins, subassemblies and fuel assemblies) followed by post-irradiation examination. Following the general trend to replace the He fuel type (High Enriched Uranium) by Leu fuel type (Low Enriched Uranium), in the light of international agreements between IAEA and the states using He fuel in their nuclear reactors, Inr Past's have been accomplished the TRIGA research reactor core full conversion on May 2006. The He fuel repatriation in US in the frame of Foreign Research Reactor Spent Nuclear Fuel Return Programme effectively started in 1999, the final stage being achieved in summer of 2008. Taking into account for the possible impact on the human and environment, in all activities associated to nuclear fuel cycle, the spent fuel or radioactive waste characteristics must be well known. Shielding calculations basic tasks consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper is a comparative study of Leu and He fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for He spent fuel, available from the last stage of the spent fuel repatriation, is presented. All the geometrical and material data related on the spent fuel shipping cask were considered according to the Nac-Lt Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface

  4. Analysis of the risk of transporting spent nuclear fuel by train

    Energy Technology Data Exchange (ETDEWEB)

    Elder, H.K.

    1981-09-01

    This report uses risk analyses to analyze the safety of transporting spent nuclear fuel for commercial rail shipping systems. The rail systems analyzed are those expected to be used in the United States when the total electricity-generating capacity by nuclear reactors is 100 GW in the late 1980s. Risk as used in this report is the product of the probability of a release of material to the environment and the consequences resulting from the release. The analysis includes risks in terms of expected fatalities from release of radioactive materials due to transportation accidents involving PWR spent fuel shipped in rail casks. The expected total risk from such shipments is 1.3 x 10/sup -4/ fatalities per year. Risk spectrums are developed for shipments of spent fuel that are 180 days and 4 years out-of-reactor. The risk from transporting spent fuel by train is much less (by 2 to 4 orders of magnitude) than the risk to society from other man-caused events such as dam failure.

  5. A method for determining the spent-fuel contribution to transport cask containment requirements

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, T.L.; Seager, K.D. [Sandia National Labs., Albuquerque, NM (United States); Rashid, Y.R.; Barrett, P.R. [ANATECH Research Corp., La Jolla, CA (United States); Malinauskas, A.P. [Oak Ridge National Lab., TN (United States); Einziger, R.E. [Pacific Northwest Lab., Richland, WA (United States); Jordan, H. [EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant; Duffey, T.A.; Sutherland, S.H. [APTEK, Inc., Colorado Springs, CO (United States); Reardon, P.C. [GRAM, Inc., Albuquerque, NM (United States)

    1992-11-01

    This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs.

  6. A method for determining the spent-fuel contribution to transport cask containment requirements

    International Nuclear Information System (INIS)

    Sanders, T.L.; Seager, K.D.; Rashid, Y.R.; Barrett, P.R.; Malinauskas, A.P.; Einziger, R.E.; Jordan, H.; Reardon, P.C.

    1992-11-01

    This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs

  7. A Historical Review of the Safe Transport of Spent Nuclear Fuel, Rev. 1

    Energy Technology Data Exchange (ETDEWEB)

    Connolly, Kevin J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pope, Ronald [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-01

    This report is a revision to M3 milestone M3FT-16OR090402028 for the former Nuclear Fuels Storage and Transportation Planning Project (NFST), “Safety Record of SNF Shipments.” The US Department of Energy (DOE) has since established the Office of Integrated Waste Management (IWM), which builds on the work begun by NFST, to develop an integrated waste management system for spent nuclear fuel (SNF), including the developm

  8. Analysis of transportation and handling system for advanced spent fuel management process

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Dong Hee; Yoon, J. S.; Park, B. S.; Ahn, S. H.; Kim, Y. H.; Jung, J. H.; Jin, J. H.; Park, G. Y.; Song, T. G. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    In this report, several devices which are used to safely transport and handle nuclear materials without scattering have been derived by analyzing the Advanced Management Process, object nuclear material and modules of process equipment and performing graphical simulation of transportation/handling by computers for the demonstration of the Advanced Spent Fuel Management Process. For verification, powder transportation vessel and handling device have been designed and manufactured. And several tests such as transporting, grappling, rotating the vessel have been performed. Also, the functional requirements of transportation/handling equipment have been analyzed based on test results and process studies. The developed functional requirements in this research will be used as the design data for the Advanced Management Process. 6 refs., 25 figs., 6 tabs. (Author)

  9. Development of Collision Accident Scenario during Nuclear Spent Fuel Maritime Transportation

    International Nuclear Information System (INIS)

    Yoo, Min; Kang, Hyun Gook

    2015-01-01

    Population density of South Korea is much higher than the other countries, and it is peninsula. Therefore, it is expected that major means of transportation of the spent fuel will be maritime transportation rather than overland transportation. Korea Maritime safety Tribunal (KMST) categorized various maritime accident, see table I. Among them, collision accident is one of the most important and complicated accident from Probabilistic Safety Analysis (PSA) point of view. We will show what will happen if the transportation ship is struck by other ship, how to calculate collision energy and probability of the branches on ship-ship collision with Event Tree Analysis (ETA) method. We selected and re-categorized maritime accident that KMST categorized for ship-ship collision analysis of spent fuel transportation ship. Event tree is constructed and collision energy distribution is derived from statistics and equation. And outer and inner hull fracture probabilities are calculated. If outer hull is broken but inner hull is fine, water will be flooded into the space between outer and inner hull. It will decrease mobility of the ship. If inner hull is fractured, water will be flooded into the ship inside. The ship has compartment structure to resist from foundering. Loss of mobility and compartment damage (ultimately it ends with sink) mechanism need to be analyzed to complete transportation ship collision event tree

  10. Accurate determination of 129I, 41Ca and 10Be long-lived radionuclides concentrations in spent resins from the nuclear industry by accelerator mass spectrometry

    International Nuclear Information System (INIS)

    Nottoli-Lepage, E.

    2013-01-01

    Radiological characterization of nuclear waste is essential for the management of storage sites. More particularly, determining the concentration of Long-Lived Radionuclides (LLRN) is fundamental for their long term management. This study focuses on the determination of three LLRN concentrations, i.e. 129 I (T 1/2 = 15.7*10 6 a), 41 Ca (T 1/2 = 9.94*10 4 a) and 10 Be (T 1/2 = 1.387*10 6 a), in ion exchange resins used for primary fluid purification in Pressurized Water Reactors (PWR). To benefit from the Accelerator Mass Spectrometry (AMS) technique allowing to measure extremely low levels of nuclide concentrations, analytical procedures including: 1) sample dissolution; 2) selective and quantitative extraction of the analyte; and, 3) analyte conditioning for AMS measurements, were developed. Applied on spent resin samples collected at a 900 MW PWR, the procedures developed for each studied LLRN allowed their quantitative recovery and their selective extraction from β-γ emitters and isobars. The concentration measurements of the LLRN of interest were then performed on the Accelerator Mass Spectrometry national facility ASTER housed by the Centre Europeen de Recherche et d'Enseignement des Geosciences de l'Environnement (CEREGE, Aix-en-Provence). 129 I, 41 Ca and 10 Be concentrations in spent resins were measured to be about 10 ng/g, 20 pg/g and 4 ng/g of dry resin, respectively. Considering 129 I and 41 Ca, the measured concentrations agree with those assessed from scaling factors established relatively to easily measured gamma emitters ( 137 Cs and 60 Co). For 10 Be, the presented results are significantly different from expected values but are in agreement with previous ICP-MS results. (author) [fr

  11. Present and future problems of radiation shielding for maritime transport of nuclear spent fuels

    International Nuclear Information System (INIS)

    Ueki, K.; Nariyama, N.; Ohashi, A.

    2000-01-01

    The transport of spent fuels with casks began in September 1999 by the exclusive spent fuel transport vessel the 'Rokuei Maru'. The casks have been transported to the reprocessing plant at Rokkasho-village in Aomori Prefecture. The 'Rokuei Maru' is approximately 100 m-length, 16.5 m-width and 3,000 gross-tons. The 20 NFT casks can be loaded into 5 holds. At the present time, the NFT casks can carry spent fuels of up to 44,000 MWD/MTU. Serpentine concrete is employed as a neutron shields in the hatch covers, the bulkheads, and the house front of the accommodations except the wheelhouse. Polyethylene covers the side walls in each hold. The neutron shielding ability of serpentine concrete and polyethylene was investigated by a shielding experiment using a 252 Cf-neutron source. The shielding experiment was analyzed with the Monte Carlo code MCNP 4B. In the near future, on-board experiment will be carried out to measure the dose-equivalent rate distributions in the 'Rokuei Maru' and the measured data and the Monte Carlo analysis of it will establish the radiation safety of the ship. (author)

  12. Development of a Computer Program (CASK) for the Analysis of Logistics and Transportation Cost of the Spent Fuels

    International Nuclear Information System (INIS)

    Cha, Jeong-Hun; Choi, Heui-Joo; Cho, Dong-Keun; Kim, Seong-Ki; Lee, Jong-Youl; Choi, Jong-Won

    2008-07-01

    The cost for the spent fuel management includes the costs for the interim storage, the transportation, and the permanent disposal of the spent fuels. The CASK(Cost and logistics Analysis program for Spent fuel transportation in Korea) program is developed to analyze logistics and transportation cost of the spent fuels. And the total amount of PWR spent fuels stored in four nuclear plant sites, a centralized interim storage facility near coast and a permanent disposal facility near the interim storage facility are considered in this program. The CASK program is developed by using Visual Basic language and coupled with an excel sheet. The excel sheet shows a change of logistics and transportation cost. Also transportation unit cost is easily changed in the excel sheet. The scopes of the report are explanation of parameters in the CASK program and a preliminary calculation. We have developed the CASK version 1.0 so far, and will update its parameters in transportation cost and transportation scenario. Also, we will incorporate it into the program which is used for the projection of spent fuels from the nuclear power plants. Finally, it is expected that the CASK program could be a part of the cost estimation tools which are under development at KAERI. And this program will be a very useful tool for the establishment of transportation scenario and transportation cost in Korean situations

  13. Development of a Computer Program (CASK) for the Analysis of Logistics and Transportation Cost of the Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Jeong-Hun; Choi, Heui-Joo; Cho, Dong-Keun; Kim, Seong-Ki; Lee, Jong-Youl; Choi, Jong-Won

    2008-07-15

    The cost for the spent fuel management includes the costs for the interim storage, the transportation, and the permanent disposal of the spent fuels. The CASK(Cost and logistics Analysis program for Spent fuel transportation in Korea) program is developed to analyze logistics and transportation cost of the spent fuels. And the total amount of PWR spent fuels stored in four nuclear plant sites, a centralized interim storage facility near coast and a permanent disposal facility near the interim storage facility are considered in this program. The CASK program is developed by using Visual Basic language and coupled with an excel sheet. The excel sheet shows a change of logistics and transportation cost. Also transportation unit cost is easily changed in the excel sheet. The scopes of the report are explanation of parameters in the CASK program and a preliminary calculation. We have developed the CASK version 1.0 so far, and will update its parameters in transportation cost and transportation scenario. Also, we will incorporate it into the program which is used for the projection of spent fuels from the nuclear power plants. Finally, it is expected that the CASK program could be a part of the cost estimation tools which are under development at KAERI. And this program will be a very useful tool for the establishment of transportation scenario and transportation cost in Korean situations.

  14. A simplified computational scheme for thermal analysis of LWR spent fuel dry storage and transportation casks

    International Nuclear Information System (INIS)

    Kim, Chang Hyun

    1997-02-01

    A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. Although a lot of computer codes and analytical models have been developed for application to the fields of thermal analysis of dry storage and/or transportation casks, some difficulties in its analysis arise from the complexity of the geometry including the rod bundles of spent fuel and the heat transfer phenomena in the cavity of cask. Particularly, if the disk-type structures such as fuel baskets and aluminium heat transfer fins are included, the thermal analysis problems in the cavity are very complex. To overcome these difficulties, cylindrical coordinate system is adopted to calculate the temperature profile of a cylindrical cask body using the multiple cylinder model as the step-1 analysis of the present study. In the step-2 analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three- dimensional conduction analysis model. The effective thermal conductivity for homogenized spent fuel assembly based on Manteufel and Todreas model is incorporated in step-2 analysis to predict the maximum fuel temperature. The presented two-step computational scheme has been performed using an existing HEATING 7.2 code and the effective thermal conductivity for the homogenized spent fuel assembly has been calculated by additional numerical analyses. Sample analyses of five cases are performed for NAC-STC including normal transportation condition to examine the applicability of the presented simplified computational scheme for thermal analysis of the large LWR spent fuel dry storage and transportation casks and heat transfer characteristics in the cavity of the cask with the disk-type structures

  15. Experience with contamination protection of spent fuel transport packages in Germany since 2000/2001

    International Nuclear Information System (INIS)

    Krinninger, H.; Bach, R.; Seidel, J.; Jung, P.

    2004-01-01

    On April 30, 1998 just a few days before the PATRAM 1998 conference at Paris, the French Nuclear Installations Safety Directorate (DSIN now DGSNR) published a press release, that during the year before some 35% of the spent fuel transports to the reprocessing plant of COGEMA at La Hague have non-fixed surface contamination in excess of the regulatory standard. A few day in advance DSIN informed in French Ministries and the competent foreign authorities of the customer countries of COGEMA. The consequences of this publication were multi-fold and perceived by the public as an act negligence of the nuclear industry. Because of concerns about additional radiation exposure to the railway workers by the unions the French Railway company SNCF suspended all transports by May 6, 1998 until implementation of corrective measures. This decision of SNF interupted also the spent fuel transports from continental Europe to the reprocessing plant of BNFL at Sellafield all performed across France to the port of Dunkirk. Furthermore on May 25, 1998 the German Federal Ministry of Environment, Nature Protection and Nuclear Safety (BMU) imposed a transport ban for shipment of spent fuel from commercial power plants and for high active waste returned from La Hague to the Gorleben site. The conditions for resumption of these transports were outlined by NMU in a 10-point programme. In response to these publications on contamination findings competent German State and Federal Authorities commissioned investigations by independent experts dealing with the identification of the causes, the proposal of counter measures, the investigation of shortcomings in the transport system in general and recommendations for retification of it

  16. Waste Management Centre (WMC) experience in providing transportation service of spent fuel radioactive sources

    International Nuclear Information System (INIS)

    Ibrahim Martibi; Mazlan Mohamad; Wilfred Paulus; Nik Marzukee Nik Ibrahim; Syed Abdul Malik Syed Zain

    2005-01-01

    MINT currently provides transportation and delivery of spent sealed sources from users back to suppliers or country of origin. This service is related to customer request and the Atomic Energy Licensing Board (AELB) policy, whereby the user is required to return back the material when it is not in use anymore. This paper discusses the service management, problems and issues faced, and improvements that can be done to increase the quality of service. The service also take into account the regulatory requirement and safety for transportation of radioactive materials as stipulated domestically and internationally. (Author)

  17. An approach for the design of closure bolts of spent fuel elements transportation packages

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Miranda, Carlos A.J.; Fainer, Gerson

    2009-01-01

    The spent fuel elements transportation packages must be designed for severe conditions including significant fire and impact loads corresponding to hypothetical accident conditions. In general, these packages have large flat lids connected to cylindrical bodies by closure bolts that can be the weak link in the containment system. The bolted closure design depends on the geometrical characteristics of the flat lid and the cylindrical body, including their flanges, on the type of the gaskets and their dimensions, and on the number, strength, and tightness of the bolts. There are well established procedures for the closure bolts design used in pressure vessels and piping. They can not be used directly in the bolts design applied to transportation packages. Prior to the use of these procedures, it is necessary consider the differences in the main loads (pressure for the pressure vessels and piping and impact loads for the transportation packages) and in the geometry (large flat lids are not used in pressure vessels and piping). So, this paper presents an approach for the design of the closure bolts of spent fuel elements transportation packages considering the impact loads and the typical geometrical configuration of the transportation packages. (author)

  18. Logistics models for the transportation of radioactive waste and spent fuel

    International Nuclear Information System (INIS)

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

    1978-03-01

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

  19. Spent fuel transport in Romania by road: An approach considering safety, risk and radiological consequences

    International Nuclear Information System (INIS)

    Vieru, G.

    2001-01-01

    The transport of high-level radioactive wastes, involving Type B packages, is a part of the safety of the Romanian waste management programme and the overall aim of this activity is to promote the safe transport of radioactive materials in Romania. The paper presents a safety case analysis of the transport of a single spent fuel CANDU bundle, using a Romanian built Type B package, from the CANDU type nuclear power plant Cernavoda to the INR Pitesti, in order to be examined within INR's hot-cells facilities. The safety assessment includes the following main aspects: (1) evaluation and analysis of available data on road traffic accidents; (2) estimation of the expected frequency for severe road accident scenarios resulting in potential radionuclide release; and (3) evaluation of the expected radiological consequences and accident risks of transport operations. (author)

  20. Analysis of burnup credit on spent fuel transport / storage casks - estimation of reactivity bias

    International Nuclear Information System (INIS)

    Mat sumura, T.; Sasahara, A.; Takei, M.; Takekawa, T.; Kagehira, K.; Nicolaou, G.; Betti, M.

    1998-01-01

    Chemical analyses of high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins were carried out. Measured data of nuclides' composition from U234 to P 242 were used for evaluation of ORIGEN-2/82 code and a nuclear fuel design code (NULIF). Critically calculations were executed for transport and storage casks for 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for axial and horizontal profiles of burnup, and historical void fraction (BWR), operational histories such as control rod insertion history, BPR insertion history and others, and calculational accuracy of ORIGEN-2/82 on nuclides' composition. This study shows that introduction of burnup credit has a large merit in criticality safety analysis of casks, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for the present reactivity bias evaluation and showed the possibility of simplifying the reactivity bias evaluation in burnup credit. (authors)

  1. Multi-purpose canisters as an alternative for storage, transportation, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Hollaway, W.R.; Rozier, R.; Nitti, D.A.; Williams, J.R.

    1993-01-01

    A study was conducted to assess the feasibility of using multi-purpose canisters to handle spent nuclear fuel throughout the Civilian Radioactive Waste Management System. Multi-purpose canisters would be sealed, metallic containers maintaining multiple spent fuel assemblies in a dry, inert environment and overpacked separately and uniquely for the various system elements of storage, transportation, and disposal. Using five implementation scenarios, the multi-purpose canister was evaluated with regard to several measures of effectiveness, including number of handlings, radiation exposure, cost, schedule and licensing considerations, and public perception. Advantages and disadvantages of the multi-purpose canister were identified relative to the current reference system within each scenario, and the scenarios were compared to determine the most effective method of implementation

  2. Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications

    International Nuclear Information System (INIS)

    Parks, C.V.; Broadhead, B.L.; Hermann, O.W.; Tang, J.S.; Cramer, S.N.; Gauthey, J.C.; Kirk, B.L.; Roussin, R.W.

    1988-07-01

    This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, and point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs

  3. Modeling of molecular and particulate transport in dry spent nuclear fuel canisters

    Science.gov (United States)

    Casella, Andrew M.

    2007-09-01

    The transportation and storage of spent nuclear fuel is one of the prominent issues facing the commercial nuclear industry today, as there is still no general consensus regarding the near- and long-term strategy for managing the back-end of the nuclear fuel cycle. The debate continues over whether the fuel cycle should remain open, in which case spent fuel will be stored at on-site reactor facilities, interim facilities, or a geologic repository; or if the fuel cycle should be closed, in which case spent fuel will be recycled. Currently, commercial spent nuclear fuel is stored at on-site reactor facilities either in pools or in dry storage containers. Increasingly, spent fuel is being moved to dry storage containers due to decreased costs relative to pools. As the number of dry spent fuel containers increases and the roles they play in the nuclear fuel cycle increase, more regulations will be enacted to ensure that they function properly. Accordingly, they will have to be carefully analyzed for normal conditions, as well as any off-normal conditions of concern. This thesis addresses the phenomena associated with one such concern; the formation of a microscopic through-wall breach in a dry storage container. Particular emphasis is placed on the depressurization of the canister, release of radioactivity, and plugging of the breach due to deposition of suspended particulates. The depressurization of a dry storage container upon the formation of a breach depends on the temperature and quantity of the fill gas, the pressure differential across the breach, and the size of the breach. The first model constructed in this thesis is capable of determining the depressurization time for a breached container as long as the associated parameters just identified allow for laminar flow through the breach. The parameters can be manipulated to quantitatively determine their effect on depressurization. This model is expanded to account for the presence of suspended particles. If

  4. Monitored Retrievable Storage (MRS) Facility and its impact on spent fuel transportation

    International Nuclear Information System (INIS)

    Joy, D.S.; Jolley, R.L.

    1986-01-01

    The Department of Energy has identified nine potential sites for a repository to permanently dispose of radioactive wastes. DOE has released several sets of maps and tables identifying expected transportation routes between nuclear reactors and repository sites. More recently, the DOE has announced three potential Monitored Retrievable Storage Facility (MRS) sites in the state of Tennessee. Obviously, if a large portion of the spent fuel is routed to Tennessee for consolidation and repackaging, there will be significant changes in the estimated routes. For typical scenarios, the number of shipments in the vicinity of the repository will be reduced. For example, with direct reactor to repository shipments, 995 highway and 262 rail shipments are expected to arrive at the repository annually. With a MRS these numbers are reduced to 201 and 30, respectively. The remaining consolidated fuel would be transported from the MRS in 22 dedicated trains (each train transporting five casks). Conversely, the MRS would result in an increase in the number of spent fuel shipments traveling through the eastern part of Tennessee. However, the operation of a MRS would significantly reduce the number of shipments through the central and western parts of the state

  5. Thermal simulations and tests in the development of a helmet transport spent fuel elements Research Reactor

    International Nuclear Information System (INIS)

    Saliba, R.; Quintana, F.; Márquez Turiello, R.; Furnari, J.C.; Pimenta Mourão, R.

    2013-01-01

    A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half-scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled. (author) [es

  6. Soviet-German cooperation in safety improvements of spent fuel transport casks

    International Nuclear Information System (INIS)

    Schulz-Forberg, B.; Zeisler, P.; Droste, B.; Kondratiev, A.; Kozlov, Ju.; Tichinov, N.

    1993-01-01

    The paper gives a survey of the Soviet (Russian)-German activities which started in 1988 with the objective of creating a long-term scientific-technical cooperation in the field of transport and storage casks for spent nuclear fuel. The first step, i.e., the step of informing each other about the state of development is done. The more complicated second phase with concerted common activities of both the Russian and German competent Authorities and industrial enterprises is intended to start the in near future. (author)

  7. Spent Nuclear Fuel Transportation: An Examination of Potential Lessons Learned From Prior Shipping Campaigns

    Energy Technology Data Exchange (ETDEWEB)

    Marsha Keister; Kathryn McBride

    2006-08-01

    The Nuclear Waste Policy Act of 1982 (NWPA), as amended, assigned the Department of Energy (DOE) responsibility for developing and managing a Federal system for the disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for accepting, transporting, and disposing of SNF and HLW at the Yucca Mountain repository in a manner that protects public health, safety, and the environment; enhances national and energy security; and merits public confidence. OCRWM faces a near-term challenge—to develop and demonstrate a transportation system that will sustain safe and efficient shipments of SNF and HLW to a repository. To better inform and improve its current planning, OCRWM has extensively reviewed plans and other documents related to past high-visibility shipping campaigns of SNF and other radioactive materials within the United States. This report summarizes the results of this review and, where appropriate, lessons learned.

  8. Analysis of the transportation logistics for spent nuclear fuel in Korea

    International Nuclear Information System (INIS)

    Lee, Hyo Jik; Ko, Won Il; Seo, Ki Seok

    2010-01-01

    As a part of the back-end fuel cycle, transportation of spent nuclear fuel (SNF) from nuclear power plants (NPP s ) to a fuel storage facility is very important in establishing a nuclear fuel cycle. In Korea, the accumulated amount of SNF in the NPP pools is troublesome since the temporary storage facilities at these NPP pools are expected to be full of SNF within ten years. Therefore, Korea cannot help but plan for the construction of an interim storage facility to solve this problem in the near future. Especially, a decision on several factors, such as where the interim storage facility should be located, how many casks a transport ship can carry at a time and how many casks are initially required, affect the configuration of the transportation system. In order to analyze the various possible candidate scenarios, we assumed four cases for the interim storage facility location, three cases for the load capacity that a transport ship can carry and two cases for the total amount of casks used for transportation. First, this study considered the currently accumulated amount of SNF in Korea, and the amount of SNF generated from NPP s until all NPP s are shut down. Then, how much SNF per year must be transported from theNPP s to an interim storage facility was calculated during an assumed transportation period. Second, 24 candidate transportation scenarios were constructed by a combination of the decision factors. To construct viable yearly transportation schedules for the selected 24 scenarios, we created a spreadsheet program named TranScenario, which was developed by using MS EXCEL. TranScenario can help schedulers input shipping routes and allocate transportation casks. Also,TranScenario provides information on the cask distribution in the NPP s and in the interim storage facility automatically, by displaying it in real time according to the shipping routes, cask types and cask numbers that the user generates. Once a yearly transportation schedule is established

  9. ANS/ENS tutorial session: Burnup credit issues in spent fuel transportation: Overview and objectives

    International Nuclear Information System (INIS)

    Sanders, T.L.

    1988-01-01

    A number of opportunities exist to increase the efficiency of the next generation of spent fuel shipping casks. Improving cask efficiency will not only reduce life cycle transportation costs, but also is consistent with maintaining public and occupational radiological risks and, more importantly, total risks (radiological and nonradiological) within the guidelines of the ''as low as reasonably achievable'' (ALARA) philosophy. Increases in cask capacities will reduce both the total number of shipments required to transport a given amount of fuel and the number of handling operations at both shipping and receiving facilities. Additional capacity increases can be achieved by implementing various design strategies based on new concepts and/or the actual characteristics of the majority of the spent fuel to be shipped in the future. For example, it has been determined that additional capacity increases can be achieved by taking credit for burnup, the reduced reactivity that results when fuel has been used to produce power in a nuclear reactor. That is, as the fuel is used the atoms of fissile material decrease, and neutron absorbers (or ''poisons'') that tend to retard the fission process are produced. 7 refs., 1 fig

  10. German Approach for the Transport of Spent Fuel Packages after Interim Storage

    International Nuclear Information System (INIS)

    Wille, Frank; Wolff, Dietmar; Droste, Bernhard; Voelzke, Holger

    2014-01-01

    In Germany the concept of dry interim storage of spent nuclear fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently (in 2012) revised 'Guidelines for dry cask storage of spent nuclear fuel and heat-generating waste' by the German Waste management Commission (ESK) which are very similar to the former RSK (reactor safety commission) guidelines. For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties which satisfy the proofs for the compliance of the safety objectives at that time. In recent years the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependent system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report including evaluation of long term behavior of components and specific operating procedures of the package. Present research and knowledge concerning the long term behavior of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behavior of aged metal and elastomeric seals under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period (aged package). Assessment methods for the material compatibility, the behavior of fuel assemblies and the aging behavior of shielding parts are issues as well. This paper describes the state

  11. TN-68 Spent Fuel Transport Cask Analytical Evaluation for Drop Events

    International Nuclear Information System (INIS)

    Shah, M.J.; Klymyshyn, Nicholas A.; Adkins, Harold E.; Koeppel, Brian J.

    2007-01-01

    The U.S. Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (10 CFR), Title 10, Part 71 (1). Both the International Atomic Energy Agency regulations for transporting radioactive materials (2, paragraph 727), and 10 CFR 71.73 require casks to be evaluated for hypothetical accident conditions, which includes a 9-meter (m) (30-ft) drop-impact event onto a flat, essentially unyielding, horizontal surface, in the most damaging orientation. This paper examines the behavior of one of the NRC certified transportation casks, the TN-68 (3), for drop-impact events. The specific area examined is the behavior of the bolted connections in the cask body and the closure lid, which are significantly loaded during the hypothetical drop-impact event. Analytical work to evaluate the NRC-certified TN-68 spent fuel transport cask (3) for a 9-m (30-ft) drop-impact event on a flat, unyielding, horizontal surface, was performed using the ANSYS (4) and LS DYNA (5) finite-element analysis codes. The models were sufficiently detailed, in the areas of bolt closure interfaces and containment boundaries, to evaluate the structural integrity of the bolted connections under 9-m (30-ft) free-drop hypothetical accident conditions, as specified in 10 CFR 71.73. Evaluation of the cask for puncture, caused by a free drop through a distance of 1-m (40-in.) onto a mild steel bar mounted on a flat, essentially unyielding, horizontal surface, required by 10 CFR 71.73, was not included in the current work, and will have to be addressed in the future. Based on the analyses performed to date, it is concluded that, even though brief separation of the flange and the lid surfaces may occur under some conditions, the seals would close at the end of the drop events, because the materials remain elastic during the duration of the event

  12. Relationship between ion transport and the failure behavior of epoxy resin coatings

    International Nuclear Information System (INIS)

    Dong, Yuhua; Zhou, Qiong

    2014-01-01

    Highlights: •An epoxy resin-Q345 system with a sandwich structure was prepared. •Cl − ions permeated into epoxy resin coating prior to K + ions. •Free volume size and PAL increased when the coating was immersed into the solution. -- Abstract: An epoxy resin coating with a sandwich structure was prepared to investigate ion transport behavior in the coating. The macro- and micro- appearance of the coating immersed in 5 wt.% KCl solutions was observed by stereomicroscopy, scanning electron microscopy equipped with an energy dispersive spectrometer. The electrochemical property of the coating was characterized by electrochemical impedance spectroscopy, and change of free volume after immersion was characterized by positron annihilation lifetime spectroscopy. The results indicated that Cl − ions permeated into the coating prior to K + ions, the free volume size and positron annihilation lifetime of the coating increased during immersion

  13. Behaviour of Aluminosilicate Inorganic Matrix Sial During and After Solidification of Radioactive Sludge And Radioactive Spent Resins and Their Mixtures

    International Nuclear Information System (INIS)

    Lichvar, P.; Rozloznik, M.; Sekely, S.

    2013-01-01

    SIAL matrix is the product of inorganic compounds by polycondensation. The inorganic SIAL matrix was subject of study from the atomic level to the level of technological application. Infrared spectroscopy measurement and X-ray analysis confirmed that SIAL matrix is mostly formed by amorphous phase. This satisfies the required criteria for disposal of immobilized radioactive waste of appropriate properties into surface repository (Slovak and Czech Republic). First of all, leaching of radioactive nuclide Cs-137 and compressive strength of the final product were found. Various remotely operated devices have been designed, manufactured and successfully used for the radioactive sludge and sludge/resins mixture solidification at the Nuclear power plant (NPP) A-1, V-2 in Jaslovske Bohunice, at Mochovce NPP (Slovak Republic), at Dukovany NPP and at Temelin NPP (Czech Republic). The SIAL matrix was approved for waste package for sludge/resins mixture by the Slovak and Czech Nuclear Regulatory Authorities. (author)

  14. BWR - Spent Fuel Transport and Storage with the TNTM9/4 and TNTM24BH Casks

    International Nuclear Information System (INIS)

    Wattez, L.; Marguerat, Y.; Hoesli, C.

    2006-01-01

    The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., the Muehleberg Nuclear Power Plant owner, is involved in this process and has elected to store its BWR spent fuel in a new high capacity dual-purpose cask, the TNeTeM24BH from the COGEMA Logistics/TRANSNUCLEAR TN TM 24 family. The Muehleberg BWR spent fuels are transported by road in a medium size shuttle transport cask and then transferred to a heavy transport/storage cask (dry transfer) in the hot cell of ZWILAG site. For that purpose, COGEMA Logistics designed and supplied: - Two shuttle casks, TN TM 9/4, mainly devoted to transport of spent fuel from Muehleberg NPP to ZWILAG. Licensed according to IAEA 1996, the TN TM 9/4 is a 40 ton transport cask, for 7 BWR high bum-up spent fuel assemblies. - A series of new high capacity dual-purpose casks, TN TM 24BH, holding 69 BWR spent fuels. Two transport campaigns took place in 2003 and 2004. For each campaign, ten TN TM 9/4 round trips are performed, and one TN TM 24BH is loaded. 5 additional TN TM 24BH are being manufactured for BKW, and the next transport campaigns are scheduled from 2006. The TN TM 24BH high capacity dual purpose cask and the TN TM 9/4 transport cask characteristics and capabilities will then be detailed. (authors)

  15. Pacific Northwest Laboratory (PNL) spent fuel transportation and handling facility models

    International Nuclear Information System (INIS)

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

    1979-09-01

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

  16. BWR-spent fuel transport and storage with the TN trademark 9/4 and TN trademark 24BH casks

    International Nuclear Information System (INIS)

    Wattez, L.; Marguerat, Y.; Hoesli, C.

    2004-01-01

    The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks in ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., as Muehleberg Nuclear Power Plant owner, is involved in this process and has selected to store its spent fuel, a new high capacity dual-purpose cask, the TN trademark 24BH. For the transport in a medium size cask, COGEMA LOGISTICS has developed a new cask, the TN trademark 9/4, to replace the NTL9 cask, which performed numerous transports of BWR spent fuel in the past decades. Licensed IAEA 1996, the TN trademark 9/4 is a 40 ton transport cask, for 7 BWR high burn-up spent fuel assemblies. The spent fuel assemblies can be transferred in the ZWILAG hot cell in the TN trademark 24BH cask. The first use of these casks took place in 2003. Ten TN trademark 9/4 transports were performed, and one TN trademark 24BH was loaded. After a brief presentation of the operational aspects, the paper will focus on the TN trademark 24BH high capacity dual purpose cask, the TN trademark 9/4 transport cask and describe in detail their characteristics and possibilities

  17. Department of Energy Spent Fuel Shipping Campaigns: Comparisons of Transportation Plans and Lessons Learned

    Energy Technology Data Exchange (ETDEWEB)

    Holm, Judith A.; Thrower, Alex W.; Antizzo, Karen

    2003-02-27

    Over the last 30 years, the U.S. Department of Energy (DOE) has successfully and safely transported shipments of spent nuclear fuel over America's highways and railroads. During that time, an exemplary safety record has been established with no identifiable fatalities, injuries, or environmental damage caused by the radioactive nature of the shipments. This paper evaluates some rail and truck shipping campaigns, planning processes, and selected transportation plans to identify lessons learned in terms of planning and programmatic activities. The intent of this evaluation is to document best practices from current processes and previous plans for DOE programs preparing or considering future plans. DOE's National Transportation Program (NTP) reviewed 13 plans, beginning with core debris shipments from Three Mile Island to current, ongoing fuel campaigns. This paper describes lessons learned in the areas of: emergency planning, planning information, security, shipment prenotification, emergency notification/response, terrorism/sabotage risk, and recovery and cleanup, as well as routing, security, carrier/driver requirements, transportation operational contingencies, tracking, inspections and safe parking.

  18. Effects of bedrock fractures on radionuclide transport near a vertical deposition hole for spent nuclear fuel

    International Nuclear Information System (INIS)

    Pulkkanen, V.-M.; Nordman, H.

    2011-12-01

    Effects of bedrock fractures on radionuclide transport near a vertical deposition hole for spent nuclear fuel are studied computationally. The studied fractures are both natural and excavation damage fractures. The emphasis is on the detailed modelling of geometry in 3D in contrast to the traditional radionuclide transport studies that often concentrate on chain decays, sorption, and precipitation at the expense of the geometry. The built computer model is used to assess the significance of components near a deposition hole for radionuclide transport and to estimate the quality of previously used modelling techniques. The results show nearly exponential decrease of radionuclide mass in the bentonite buffer when the release route is a thin natural fracture. The results also imply that size is the most important property of the tunnel section for radionuclide transport. In addition, the results demonstrate that the boundary layer theory can be used to approximate the release of radionuclides with certain accuracy and that a thin fracture in rock can be modelled, at least to a certain limit, by using a fracture with wider aperture but with same flow rate as the thin fracture. (orig.)

  19. Adsorption of Rh(III) complexes from chloride solutions obtained by leaching chlorinated spent automotive catalysts on ion-exchange resin Diaion WA21J

    International Nuclear Information System (INIS)

    Shen Shaobo; Pan Tonglin; Liu Xinqiang; Yuan Lei; Wang Jinchao; Zhang Yongjian; Guo Zhanchen

    2010-01-01

    It was found that Rh, Pd and Pt contained in the spent ceramic automotive catalysts could be effectively extracted by dry chlorination with chlorine. In order to concentrate Rh(III) ions contained in the chloride solutions obtained, thermodynamic and kinetics studies for adsorption of Rh(III) complexes from the chloride solutions on an anionic exchange resin Diaion WA21J were carried out. Rh, Pd, Pt, Al, Fe, Si, Zn and Pb from the chloride solution could be adsorbed on the resin. The distribution coefficients (K d ) of Rh(III) decreased with the increase in initial Rh(III) concentration or in adsorption temperature. The isothermal adsorption of Rh(III) was found to fit Langmuir, Freundlich and Dubinin-Kaganer-Radushkevich models under the adsorption conditions. The maximum monolayer adsorption capacities Q max based on Langmuir adsorption isotherms were 6.39, 6.61 and 5.81 mg/g for temperatures 18, 28 and 40 deg. C, respectively. The apparent adsorption energy of Rh was about -7.6 kJ/mol and thus Rh(III) adsorption was a physical type. The experimental data obtained could be better simulated by pseudo-first-order kinetic model and the activation energy obtained was 6.54 J/mol. The adsorption rate of Rh(III) was controlled by intraparticle diffusion in most of time of adsorption process.

  20. Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications. [Radiation dose rates from shielded spent fuels and high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Parks, C.V.; Broadhead, B.L.; Hermann, O.W.; Tang, J.S.; Cramer, S.N.; Gauthey, J.C.; Kirk, B.L.; Roussin, R.W.

    1988-07-01

    This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, and point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs.

  1. Transport of spent nuclear fuel from the High Flux Beam Reactor

    International Nuclear Information System (INIS)

    Holland, Michael; Carelli, Joseph; Shelton, Thomas

    1997-01-01

    The shipment of more than 1000 elements of spent nuclear fuel (SNF) from the Department of Energy's Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR) to the Department's Savannah River Site (SRS) for long term interim storage required overcoming several significant obstacles. The project management team was comprised of DOE, BNL and NAC International personnel. This achievement involved coordinating the efforts of numerous government and contractor organizations such as the U.S. Coast Guard, the U.S. Nuclear Regulatory Commission, state and local governments, marine and motor carriers, and carrier inspectors. Unique experience was gained during development and execution of the project in the following areas: dry transfer of SNF to shipping casks; inter-modal transfers; logistics; cask licensing by the Nuclear Regulatory Commission (NRC); compliance with environmental regulations; transportation plan development, and stakeholder outreach and coordination

  2. Thick nickel plating of spent fuel transport and storage casks CASTOR and POLLUX

    International Nuclear Information System (INIS)

    Wilbuer, K.

    1991-01-01

    Spent fuel elements have to be safely handled in containers for transport and storage. These large casks (100-120 t) are made by various firms according to the specifications given by the nuclear plant operator. For shielding and protection of the hazardous material, the casks' inner surface is coated with a nickel plating about 3000 μm thick. The product and the production process are subject to very stringent requirements, due to the hazardous potential of the material to be shipped or stored. Therefore, both the extremely high quality standards to be met by the nickel plating and the dimensions and capability of the plating plant required for the process are problems that cannot be solved by a usual commercial plating plant. The new concept and process that had to be established are explained in the paper. (orig./MM) [de

  3. Calculation of health risks from spent-nuclear-fuel transportation accidents

    International Nuclear Information System (INIS)

    Chen, S.Y.; Yuan, Y.C.

    1988-01-01

    Models developed to analyze potential radiological health risks from various accident scenarios during transportation of spent nuclear fuels are described. The models are designed both for detailed route-specific risk analyses and for use in conducting overall risk analyses for route selection and related decision-making activities. The radiological risks calculated include individual dose commitments, collective dose commitments, and long-term (100-year) environmental dose commitments to a population following release of radioactivity. To facilitate route-specific analysis, a state-levle database was developed and incorporated into the model. Route-specific analysis is demonstrated by the calculation of radiological risks resulting from various accident scenarios, as postulated by the recent US Nuclear Regulatory Commission Modal Study, for four representative states selected from various regions of the US

  4. ANSI N14.5 source term licensing of spent-fuel transport cask containment

    International Nuclear Information System (INIS)

    Seager, K.D.; Reardon, P.C.; James, R.J.; Foadian, H.; Rashid, Y.R.

    1993-01-01

    American National Standards Institute (ANSI) standard N14.5 states that ''compliance with package containment requirements shall be demonstrated either by determination of the radioactive contents release rate or by measurement of a tracer material leakage rate.'' The maximum permissible leakage rate from the transport cask is equal to the maximum permissible release rate divided by the time-averaged volumetric concentration of suspended radioactivity within the cask. The development of source term methodologies at Sandia National Laboratories (SNL) provides a means to determine the releasable radionuclide concentrations within spent-fuel transport casks by estimating the probability of cladding breach, quantifying the amount of radioactive material released into the cask interior from the breached fuel rods, and quantifying the amount of radioactive material within the cask due to other sources. These methodologies are implemented in the Source Term Analyses for Containment Evaluations (STACE) software. In this paper, the maximum permissible leakage rates for the normal and hypothetical accident transport conditions defined by 10 CFR 71 are estimated using STACE for a given cask design, fuel assembly, and initial conditions. These calculations are based on defensible analysis techniques that credit multiple release barriers, including the cladding and the internal cask walls

  5. Hypothetical Case and Scenario Description for International Transportation of Spent Nuclear Fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Adam David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, Douglas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Jones, Katherine A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kalinina, Elena Arkadievna [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cohn, Brian [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Thomas, Maikael A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Parks, Mancel Jordan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Parks, Ethan Rutledge [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mohagheghi, Amir H. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-12-01

    To support more rigorous analysis on global security issues at Sandia National Laboratories (SNL), there is a need to develop realistic data sets without using "real" data or identifying "real" vulnerabilities, hazards or geopolitically embarrassing shortcomings. In response, an interdisciplinary team led by subject matter experts in SNL's Center for Global Security and Cooperation (CGSC) developed a hypothetical case description. This hypothetical case description assigns various attributes related to international SNF transportation that are representative, illustrative and indicative of "real" characteristics of "real" countries. There is no intent to identify any particular country and any similarity with specific real-world events is purely coincidental. To support the goal of this report to provide a case description (and set of scenarios of concern) for international SNF transportation inclusive of as much "real-world" complexity as possible -- without crossing over into politically sensitive or classified information -- this SAND report provides a subject matter expert-validated (and detailed) description of both technical and political influences on the international transportation of spent nuclear fuel. [PAGE INTENTIONALLY LEFT BLANK

  6. Transport of HIFAR spent fuel from Lucas Heights Research Establishment to the United Kingdom for reprocessing. Public Environmental Report

    International Nuclear Information System (INIS)

    1995-01-01

    The normal operations of HIFAR produce thirty-eight spent fuel elements annually. Since 1958, when operations began, 1,660 spent fuel elements have been accumulated and are stored in ANSTO's engineered interim storage facilities at Lucas Heights. In the light of the limited size of these storage facilities and following the Research Reactor Review (1993) and an Inter-Agency Review, the Commonwealth Government announced its decision to reduce the number of spent fuel elements stored at the site. Therefore, ANSTO has been authorised to negotiate the terms for shipment of spent fuel elements of United Kingdom (UK) origin to the Dounreay reprocessing plant in Scotland. This Public Environment Report, prepared under the Environment Protection (Impact of Proposals) Act 1974, describes the potential impacts and risks of a proposed initial shipment of 120 spent fuel elements to the Dounreay reprocessing plant. It describes the intended packaging and transport procedures and considers possible alternative methods of dealing with the continued production of spent fuel rods and the limited storage capacity at LHRL. The exhaustive analysis of every phase of operations involved in the shipping of a cask of spent HIFAR fuel elements from Lucas Heights to Dounreay, for reprocessing, has shown that there are no significant environmental or public health impacts from such a shipment conducted in accordance with standard, internationally established procedures. 18 refs., 12 tabs., 2 figs

  7. Regional dual-purpose cask for the storage and transport of research reactor spent fuel

    International Nuclear Information System (INIS)

    Neto, Miguel Mattar; Mourao, Rogerio Pimenta

    2002-01-01

    Taking into account that the deadline set for the American program of taking back foreign research reactor spent fuel containing U.S.-supplied enriched uranium - the year 2006, the five Latin American countries operating this type of reactor - Argentina, Brazil, Chile, Peru and Mexico - decided to launch an IAEA-sponsored project aiming at establishing local expertise in managing this material. Among the alternatives for an extended storage of the disused elements, the use of a dual purpose cask for both storage and transport is being seriously considered, due to its appealing advantages: expansion of the plant storage capacity without the burden of costly modifications of the reactor building, flexibility, in that the used fuel can be stored in situ or in other facilities outside the reactor site and the preparation of the elements for the future transportation to the final repository. At the present stage, the cask conceptual design is being developed at the Brazilian participating institutes - CDTN and IPEN. The basic idea is to work on a concept which meets the needs and particularities of each country, in terms of fuel type and dimensions, reactor building handling and transport capabilities, expected spent fuel production, etc, and also be approved by the licensing authorities of all countries involved. The preliminary concept is of a cylindrical cask with an internal cavity, a basket to hold the fuel elements and external shock absorbers. The main body is a sturdy structure with external surfaces of stainless steel and lead filling, which provides the necessary shielding. A double lid system with gaskets and inspection ports guarantees containment and control over any possible gas leakage. Due to the different fuel types used in Latin American research reactors - both MTR and TRIGA fuels are used - and to allow for the storage and transportation of processed fuel, different internal basket designs will be developed. The external shock absorbers are filled

  8. Spent Nuclear Fuel Transportation: An Examination of Potential Lessons Learned From Prior Shipping Campaigns

    International Nuclear Information System (INIS)

    M. Keister; K, McBride

    2006-01-01

    The Nuclear Waste Policy Act of 1982 (NWPA), as amended, assigned the Department of Energy (DOE) responsibility for developing and managing a Federal system for the disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for accepting, transporting, and disposing of SNF and HLW at the Yucca Mountain repository (if licensed) in a manner that protects public health, safety, and the environment; enhances national and energy security; and merits public confidence. OCRWM faces a near-term challenge--to develop and demonstrate a transportation system that will sustain safe and efficient shipments of SNF and HLW to a repository. To better inform and improve its current planning, OCRWM has extensively reviewed plans and other documents related to past high-visibility shipping campaigns of SNF and other radioactive materials within the United States. This report summarizes the results of this review and, where appropriate, lessons learned. The objective of this lessons learned study was to identify successful, best-in-class trends and commonalities from past shipping campaigns, which OCRWM could consider when planning for the development and operation of a repository transportation system. Note: this paper is for analytical and discussion purposes only, and is not an endorsement of, or commitment by, OCRWM to follow any of the comments or trends. If OCRWM elects to make such commitments at a future time, they will be appropriately documented in formal programmatic policy statements, plans and procedures. Reviewers examined an extensive study completed in 2003 by DOE's National Transportation Program (NTP), Office of Environmental Management (EM), as well as plans and documents related to SNF shipments since issuance of the NTP report. OCRWM examined specific planning, business, institutional and operating practices that have been identified by DOE, its transportation contractors

  9. Spent fuel storage and transport cask decontamination and modification. An overview of management requirements and applications based on practical experience

    International Nuclear Information System (INIS)

    1999-04-01

    A large increase in the number of casks required for transport and/or storage of spent fuel is forecast into the next century. The principal requirement will be for increased number of storage and dual purpose (transport/storage) casks for interim storage of spent fuel prior to reprocessing or permanent disposal in both on-site and off-site storage facilities. Through contact with radioactive materials spent fuel casks will be contaminated on both internal and external surfaces. In broad terms, cask contamination management can be defined by three components: minimisation, prevention and decontamination. This publication is a compilation of international experience with cask contamination problems and decontamination practices. The objective is to present current knowledge and experience as well as developments, trends and potential for new applications in this field. Furthermore, the report may assist in new design or modification of existing casks, cask handling systems and decontamination equipment

  10. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    International Nuclear Information System (INIS)

    Yuan, Y.C.; Chen, S.Y.; LePoire, D.J.

    1993-02-01

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors

  11. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Y.C. [Square Y, Orchard Park, NY (United States); Chen, S.Y.; LePoire, D.J. [Argonne National Lab., IL (United States). Environmental Assessment and Information Sciences Div.; Rothman, R. [USDOE Idaho Field Office, Idaho Falls, ID (United States)

    1993-02-01

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors.

  12. CHARACTERISTICS OF NEXT-GENERATION SPENT NUCLEAR FUEL (SNF) TRANSPORT AND STORAGE CASKS

    Energy Technology Data Exchange (ETDEWEB)

    Haire, M.J.; Forsberg, C.W.; Matveev, V.Z.; Shapovalov, V.I.

    2004-10-03

    The design of spent nuclear fuel (SNF) casks used in the present SNF disposition systems has evolved from early concepts about the nuclear fuel cycle. The reality today is much different from that envisioned by early nuclear scientists. Most SNF is placed in pool storage, awaiting reprocessing (as in Russia) or disposal at a geologic SNF repository (as in the United States). Very little transport of SNF occurs. This paper examines the requirements for SNF casks from today's perspective and attempts to answer this question: What type of SNF cask would be produced if we were to start over and design SNF casks based on today's requirements? The characteristics for a next-generation SNF cask system are examined and are found to be essentially the same in Russia and the United States. It appears that the new depleted uranium dioxide (DUO2)-steel cermet material will enable these requirements to be met. Depleted uranium (DU) is uranium in which a portion of the 235U isotope has been removed during a uranium enrichment process. The DUO2-steel cermet material is described. The United States and Russia are cooperating toward the development of a next-generation, dual-purpose, storage and transport SNF system.

  13. Routing and scheduling of hazardous materials shipments: algorithmic approaches to managing spent nuclear fuel transport

    International Nuclear Information System (INIS)

    Cox, R.G.

    1984-01-01

    Much controversy surrounds government regulation of routing and scheduling of Hazardous Materials Transportation (HMT). Increases in operating costs must be balanced against expected benefits from local HMT bans and curfews when promulgating or preempting HMT regulations. Algorithmic approaches for evaluating HMT routing and scheduling regulatory policy are described. A review of current US HMT regulatory policy is presented to provide a context for the analysis. Next, a multiobjective shortest path algorithm to find the set of efficient routes under conflicting objectives is presented. This algorithm generates all efficient routes under any partial ordering in a single pass through the network. Also, scheduling algorithms are presented to estimate the travel time delay due to HMT curfews along a route. Algorithms are presented assuming either deterministic or stochastic travel times between curfew cities and also possible rerouting to avoid such cities. These algorithms are applied to the case study of US highway transport of spent nuclear fuel from reactors to permanent repositories. Two data sets were used. One data set included the US Interstate Highway System (IHS) network with reactor locations, possible repository sites, and 150 heavily populated areas (HPAs). The other data set contained estimates of the population residing with 0.5 miles of the IHS and the Eastern US. Curfew delay is dramatically reduced by optimally scheduling departure times unless inter-HPA travel times are highly uncertain. Rerouting shipments to avoid HPAs is a less efficient approach to reducing delay

  14. Transport - mountain or molehill? The transport of spent fuel - How a minor incident became a major issue

    International Nuclear Information System (INIS)

    Daifuku, Karen

    1999-01-01

    On May 6, 1998 the national newspaper 'Liberation' published a story revealing that some of the spent fuel casks shipped from Electricite de France's NPPs would arrive at Cogema's rail terminal at Valognes with some surface contamination at levels above the authorized limits. Even though the contamination never represented a health hazard to either the population or to the operatives due to its very low levels, the story turned into a communication crisis for the nuclear industry. Despite the best laid out plans, events can take one by surprise as illustrated in the transport of spent fuel issue. So, it is now a question of anticipating better, putting in place an improved alert system, monitoring events and being able to make the proper analysis in order to prevent benign situations from getting out of hand. Adaptation to an ever evolving environment is another important factor. Knowing your opposition, not just those that are familiar like clearly identified foes such as environmental groups, but also groups or individuals that can be in the mainstream of society, including members of the government. Experience has showed that openness is the key to a relationship of trust at all levels, with the general public, politicians and the media. It does not mean total disclosure at all times because of the risk of overwhelming one's target audience with information without them being able to either evaluate its importance or sometimes fully comprehend it if it is complex. As for the media, it is crucial to built a strong relationship with journalists. It is a painstaking operation not only necessary but crucial. Following-up is a must as it permits to make sure there are no misunderstandings and helps put information into the proper perspective

  15. A comparison of radiation doses and risks between spent fuel transport/storage and selected non-nuclear activities

    International Nuclear Information System (INIS)

    Pennington, C.W.

    2003-01-01

    Spent fuel transport and storage have achieved an exemplary safety record over four decades within both the United States (US) and the global community at large. This paper offers an assessment demonstrating the safety of spent fuel transport and storage packagings relative to currently accepted but unregulated non-nuclear activities and practices within society. Over the last quarter of a century, several spent fuel transport and storage packaging test programmes have produced data that allow calculation of potential releases and population doses resulting from a terrorist attack. The US Department of Energy (DOE) has used this information to develop projected worst-case, low probability population exposures as part of the Final Environmental Impact Statement (FEIS) for the Yucca Mountain repository. The paper discusses potential population exposures from these packagings based on analysis and testing under beyond-design-basis (BDB) events, including missile attacks, and then defines and defends an acceptance criterion for the bounding outcomes of these events, based upon current accepted activities within society that produce high radiation doses to the general public. These activities involve unregulated technologies and practices within society that yield population doses significantly exceeding those that would result from such hypothetical and highly improbable events as a terrorist missile attack on a spent fuel transport or storage packaging. In particular, technologically enhanced natural radiation (TENR) exposures from building materials, farming, and masonry construction are highlighted. Recent landmark work by the US National Academy of Sciences (NAS) and by the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) are cited in support of this assessment, along with work from the US Environmental Protection Agency (EPA). From this compelling evidence, it is concluded that spent fuel transport and storage represent a low

  16. Transportation and storage optimization of spent nuclear fuel in the United States

    International Nuclear Information System (INIS)

    Gupta, S.; Boryczka, M.K.; Collyer, P.L.

    1990-01-01

    In 1987, the U.S. Congress amended the Nuclear Waste Policy Act, establishing an independent body known as the Monitored Retrievable Storage Review Commission. The mandate of the Commission was to review DOE's assessment of the advantages and disadvantages of the Monitored Retrievable Storage (MRS) facility and to evaluate the need for an MRS. The MRS Review Commission asked ICF Technology Incorporated to develop a systems integration and transportation analysis model that would aid their evaluation for Congress. The resulting model (Model for Analysis of System Risk And Cost - MARC) was adapted from an earlier model (TRICAM) written for the DOE. MARC was made available to the Commission to calculate the costs and risks (both radiological and non-radiological) of alternative spent fuel storage configurations within the Federal Waste Management System. Two of the over 30 MARC scenarios run by the MRS Commission are used to demonstrate the use of MARC in evaluating alternatives in terms of system costs and risks. These two cases are initially run in a cost minimizing mode and then in a risk minimizing mode in order to compare the difference in the value of risk for each system configuration. This example demonstrates the kind of insights MARC can yield and its utility in policy-making. 4 refs., 2 figs., 2 tabs

  17. Comparison and Analysis of Regulatory and Derived Requirements for Certain DOE Spent Nuclear Fuel Shipments; Lessons Learned for Future Spent Fuel Transportation Campaigns

    International Nuclear Information System (INIS)

    Kramer, George L.; Fawcett, Rick L.; Rieke, Philip C.

    2003-01-01

    Radioactive materials transportation is stringently regulated by the Department of Transportation and the Nuclear Regulatory Commission to protect the public and the environment. As a Federal agency, however, the U.S. Department of Energy (DOE) must seek State, Tribal and local input on safety issues for certain transportation activities. This interaction has invariably resulted in the imposition of extra-regulatory requirements, greatly increasing transportation costs and delaying schedules while not significantly enhancing the level of safety. This paper discusses the results an analysis of the regulatory and negotiated requirements established for a July 1998 shipment of spent nuclear fuel from foreign countries through the west coast to the Idaho National Engineering and Environmental Laboratory (INEEL). Staff from the INEEL Nuclear Materials Engineering and Disposition Department undertook the analysis in partnership with HMTC, to discover if there were instances where requirements derived from stakeholder interactions duplicate, contradict, or otherwise overlap with regulatory requirements. The study exhaustively lists and classifies applicable Department of Transportation (DOT) and Nuclear Regulatory Commission (NRC) regulations. These are then compared with a similarly classified list of requirements from the Environmental Impact Statements (EIS) and those developed during stakeholder negotiations. Comparison and analysis reveals numerous attempts to reduce transportation risk by imposing more stringent safety measures than those required by DOT and NRC. These usually took the form of additional inspection, notification and planning requirements. There are also many instances of overlap with, and duplication of regulations. Participants will gain a greater appreciation for the need to understand the risk-oriented basis of the radioactive materials regulations and their effectiveness in ensuring safety when negotiating extra-regulatory requirements

  18. Summary of the spent nuclear fuel transportation workshop for state officials, October 22, 1985, Springfield, Illinois. Final report

    International Nuclear Information System (INIS)

    1985-01-01

    On October 22, 1985, thirty-two state officials representing eighteen states attended a workshop on the transportation of spent nuclear fuel hosted by the Illinois Department of Nuclear Safety and co-sponsored by the US Department of Energy. The letter of invitation is attached to this report along with the agenda for the workshop. Also attached is the attendance list. The purposes of the workshop were to explain Illinois' inspection, escort and emergency preparedness program for spent nuclear fuel to officials from other states and to discuss the needs of the future national system when DOE begins shipping spent fuel to either a repository or a monitored retrievable storage facility. Summaries of the presentations and the reports from the discussion groups are presented

  19. Spent fuels program

    International Nuclear Information System (INIS)

    Shappert, L.B.

    1983-01-01

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

  20. Development of dual-purpose metal cask for interim storage of spent nuclear fuel (3). A new type of durable neutron shielding resin

    International Nuclear Information System (INIS)

    Kamoshida, Mamoru; Nishi, Takashi; Iga, Kiminori; Shimizu, Masashi; Kashiwakura, Jun; Hayashi, Makoto

    2003-01-01

    Hitachi Ltd. has been developing a new neutron shielding resin for the dual-purpose metal cask. The newly developed resin is composed of a thermo setting epoxy and magnesium hydroxide. Highly durable resin can be obtained by combining base polymer having dense cross linkage and fire retardant with high dehydration temperature. Estimated weight loss of the resin during storage period is less than 1%. Fabrication process of shielding unit suitable for the new material is also developed. In the new process, a resin block constituted of cured resin and heat transfer fin is manufactured and fitted to the cask. This process was verified by fabricating about 200 resin blocks for real size mock-up of Hitachi's metal cask. (author)

  1. Market driven strategy for acquisition of waste acceptance and transportation services for commercial spent fuel in the United States

    International Nuclear Information System (INIS)

    Lemeshewky, W.; Macaluso, C.; Smith, P.; Teer, B.

    1998-05-01

    The Department of Energy has the responsibility for the shipment of spent nuclear fuel (SNF) from commercial reactors to a Federal facility for storage and/or disposal. DOE has developed a strategy for a market driven approach for the acquisition of transportation services and equipment which will maximize the participation of private industry. To implement this strategy, DOE is planning to issue a Request for Proposal (RFP) for the provision of the required services and equipment to accept SNF from the utilities and transport the SNF to a Federal facility. The paper discusses this strategy and describes the RFP

  2. A report on the transport of MTR-type spent fuel assemblies of the Philippine Research Reactor (PRR-1)

    International Nuclear Information System (INIS)

    Yoshisaki, Magno B.; Leopando, Leonardo S.

    1999-03-01

    Fifty one (51) fuel assemblies of mixed enrichment from the Philippine Research Reactor (PRR-1), consisting of 50 spent and 1 fresh, were shipped to the United States last 14 March 1999 under the U.S. Return of Foreign Research Reactor (FRR) fuel policy. The shipment was in line with the U.S. initiative to implement its Record of Decision (ROD) which took effect on 13 May 1996 to accept and manage all FRR uranium fuel of U.S. origin and enriched in the United States. The shipment program would last10 years, ending midnight of 13 May 2006. The ROD provided a 3 year extension period within which to accept FRR spent nuclear fuel (SNF) withdrawn from reactors after 2006. The U.S. policy gave priority to the NPT significance of high enriched U, as the prime target of the return of FRR policy. Classified as a developing country, the Philippines, through the PNRI, signed a contract with the U.S. Department of Energy for the cost-free shipment of PRR-1 spent fuel to the United States. Spent fuel loading and transport operations to the port area lasted seven (7) days, from 8 to 14 March 1999. (Author)

  3. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    International Nuclear Information System (INIS)

    Yuan, Y.C.; Chen, S.Y.; Biwer, B.M.; LePoire, D.J.

    1995-11-01

    This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows trademark environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident

  4. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Y.C. [Square Y Consultants, Orchard Park, NY (US); Chen, S.Y.; Biwer, B.M.; LePoire, D.J. [Argonne National Lab., IL (US)

    1995-11-01

    This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows{trademark} environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident.

  5. Proceedings of a workshop on the use of burnup credit in spent fuel transport casks

    International Nuclear Information System (INIS)

    Sanders, T.L.

    1989-10-01

    The Department of Energy sponsored a workshop on the use of burnup credit in the criticality design of spent fuel shipping casks on February 21 and 22, 1988. Twenty-five different presentations on many related topics were conducted, including the effects of burnup credit on the design and operation of spent fuel storage pools, casks and modules, and shipping casks; analysis and physics issues related to burnup credit; regulatory issues and criticality safety; economic incentives and risks associated with burnup credit; and methods for verifying spent fuel characteristics. An abbreviated version of the DOE workshop was repeated as a special session at the November 1988 American Nuclear Society Meeting in Washington, DC. Each of the invited speakers prepared detailed papers on his or her respective topic. The individual papers have been cataloged separately

  6. Assessment of the safety of spent fuel transportation in urban environs

    International Nuclear Information System (INIS)

    Sandoval, R.P.; Weber, J.P.; Levine, H.S.

    1983-06-01

    The results of a program to provide an experimental data base for estimating the radiological consequences from a hypothetical sabotage attack on a light-water-reactor spent fuel shipping cask in a densely populated area are presented. The results of subscale and full-scale experiments in conjunction with an analytical modeling study are described. The experimental data were used as input to a reactor-safety consequence model to predict radiological health consequences resulting from a hypothetical sabotage attack on a spent-fuel shipping cask in the Manhattan borough of New York City. The results of these calculations are presented

  7. Free volumes and gas transport in polymers: amine-modified epoxy resins as a case study.

    Science.gov (United States)

    Patil, Pushkar N; Roilo, David; Brusa, Roberto S; Miotello, Antonio; Aghion, Stefano; Ferragut, Rafael; Checchetto, Riccardo

    2016-02-07

    The CO2 transport process was studied in a series of amine-modified epoxy resins having different cross-linking densities but the same chemical environment for the penetrant molecules. Positron Annihilation Lifetime Spectroscopy (PALS) was used to monitor the free volume structure of the samples and experimentally evaluate their fractional free volume fh(T) and its temperature evolution. The analysis of the free volume hole size distribution showed that all the holes have a size large enough to accommodate the penetrant molecules at temperatures T above the glass transition temperature Tg. The measured gas diffusion constants at T > Tg have been reproduced in the framework of the free volume theory of diffusion using a novel procedure based on the use of fh(T) as an input experimental parameter.

  8. Determination of {sup 55}Fe and {sup 63}Ni in spent ion-exchange resins from the IEA-R1 reactor (IPEN-CNEN/SP)

    Energy Technology Data Exchange (ETDEWEB)

    Taddei, Maria Helena T.; Ferreira, Marcelo T., E-mail: mhtaddei@cnen.gov.b, E-mail: ferreira@cnen.gov.b [Brazilian Nuclear Energy Commission (LAPOC/CNEN-MG), Pocos de Caldas, MG (Brazil). Lab. of Pocos de Caldas; Vicente, Roberto; Marumo, Julio T.; Terremoto, Luis A. Albiac, E-mail: jtmarumo@ipen.b, E-mail: rvicente@ipen.b, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    Iron and nickel are used in a wide range of metallic alloys employed in reactor core structures and their corresponding activation products are often encountered in reactor-derived solid low-level wastes and effluents. Both {sup 63} Ni and {sup 55} Fe are neutron activation products. {sup 63} Ni is produced by a neutron-gamma reaction of {sup 62} Ni and by a neutron-proton reaction of-{sup 63}Cu. Iron-55 is produced by neutron activation of two major stable iron isotopes: neutron-gamma reaction of {sup 54} Fe and neutron-to-neutron reaction of {sup 56}Fe. Nickel-63 is a pure beta emitter with maximum energy of 66.95 keV and half-life of 100.1 years, and {sup 55}Fe, with a half life of 2.7 years, decays via electron capture to stable {sup 55}Mn with the emission of Auger electrons and 5.89 keV X-rays. {sup 63} Ni is an important radionuclide in safety analysis of nuclear waste repositories and {sup 55} Fe is an important contributor to the radioactivity of nuclear waste in the first few years of storage. In this work, liquid scintillation counting was used for the determination of both radionuclides in spent ion-exchange resins taken from the water retreatment system of the IEA-R1 reactor and stored at the Radioactive Waste Management Laboratory of the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Since nuclear waste usually contains many beta emitters, such as {sup 3}H, {sup 14}C, {sup 60}Co, {sup 90}Sr, and {sup 137}Cs, a highly efficient chemical separation method is required to isolate {sup 55}Fe and {sup 63}Ni from all other radionuclides. Hydroxide precipitation was used to separate {sup 55} Fe and {sup 63}Ni from the waste matrix and ion exchange chromatography was used to separate {sup 55}Fe and {sup 63}Ni from each other as well as from interfering radionuclides. The chemical yield was above 90% for {sup 63}Ni and above 60% for {sup 55}Fe. The results of activity measurements correlated well with the concentration of {sup 60}Co in this waste

  9. Time spent by Brazilian students in different modes of transport going to school: changes over a decade (2001-2011

    Directory of Open Access Journals (Sweden)

    Kelly Samara Silva

    2014-11-01

    Full Text Available To examine changes in the time spent in each mode of transportation used for going to school by gender and age among adolescents from Santa Catarina State, Brazil. Two school-based surveys were performed in 2001 (N = 5,028 and 2011 (N = 6,529 in high school students (15-19 years old. The mode of transportation (on foot; by bicycle; by bus; car/motorcycle and the time spent for commuting to school were assessed. Active commuting increased for short trips in both genders (male: 25.1% to 36.7%; female: 18.8% to 29.2% and in all ages (15-16 years: 21% to 32.7%; 17-19 years: 21.9% to 32.4%, and declined for longer trips in males (30.5% to 21.9% and in 15-16 years old students (25.7% to 34.7%. Car/motorcycle use has doubled for short trips in males (38.1% to 65.9% and in 17-19 years old students (37.7% to 62.7%, while the use of buses remained stable in both genders. Our findings contribute to discussions on public policy focusing on the design of safe environments to promote active commuting to schools, particularly to decrease the use of motorized transport for short trips.

  10. A risk-informed basis for establishing non-fixed surface contamination limits for spent fuel transportation casks

    International Nuclear Information System (INIS)

    Rawl, R.R.; Eckerman, K.F.; Bogard, J.S.; Cook, J.R.

    2004-01-01

    The current limits for non-fixed contamination on packages used to transport radioactive materials were introduced in the 1961 edition of the International Atomic Energy Agency (IAEA) transport regulations and were based on radiation protection guidance and practices in use at that time. The limits were based on exposure scenarios leading to intakes of radionuclides by inhalation and external irradiation of the hands. These considerations are collectively referred to as the Fairbairn model. Although formulated over 40 years ago, the model remains unchanged and is still the basis of current regulatory-derived limits on package non-fixed surface contamination. There can also be doses that while not resulting directly from the contamination, are strongly influenced by and attributable to transport regulatory requirements for contamination control. For example, actions necessary to comply with the current derived limits for light-water-reactor (LWR) spent nuclear fuel (SNF) casks can result in significant external doses to workers. This is due to the relatively high radiation levels around the loaded casks, where workers must function during the measurement of contamination levels and while decontaminating the cask. In order to optimize the total dose received due to compliance with cask contamination levels, it is necessary to take into account all the doses that vary as a result of the regulatory limit. Limits for non-fixed surface contamination on spent fuel casks should be established by using a model that considers and optimizes the appropriate exposure scenarios both in the workplace and in the public environment. A risk-informed approach is needed to ensure optimal use of personnel and material resources for SNF-based packaging operations. This paper is a summary of a study sponsored by the US Nuclear Regulatory Commission and performed by Oak Ridge National Laboratory that examined the dose implications for removable surface contamination limits on spent fuel

  11. CASTOR(r) and CONSTOR(r) type transport and storage casks for spent fuel and high active waste

    International Nuclear Information System (INIS)

    Kuehne, B.; Sowa, W.

    2002-01-01

    The German company GNB has developed, tested, licensed, fabricated, loaded, transported and stored a large number of casks for spent fuel and high-level waste. CASTOR(r) casks are used at 18 sites on three continents. Spent fuel assemblies of the types PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high active waste (HAW) containers are stored in these kinds of casks. More than 600 CASTOR(r) casks have been loaded for long-term storage. The two decades of storage have shown that the basic requirements, which are safe confinement, criticality safety, sufficient shielding and appropriate heat transfer have been fulfilled in each case. There is no indication that problems will arise in the future. Of course, the experience of 20 years has resulted in improvements of the cask design. One basic improvement is GNB's development since the mid 1990s of a sandwich cask design using heavy concrete and steel as basic materials, for economical and technical reasons. This CONSTOR(r) cask concept also fulfils all design criteria for transport and storage given by the IAEA recommendations and national authorities. By May 2002 40 CONSTOR(r) casks had been delivered and 15 had been successfully loaded and stored. In this paper the different types of casks are presented. Experiences gained during the large number of cask loadings and more than 4000 cask-years of storage will be summarised. The presentation of recent and future development shows the optimisation potential of the CASTOR(r) and CONSTOR(r) cask families for safe and economical management of spent fuel. (author)

  12. Regulatory status of burnup credit for dry storage and transport of spent nuclear fuel in the United States

    International Nuclear Information System (INIS)

    Carlson, D.E.

    2001-01-01

    During 1999, the Spent Fuel Project Office of the U.S. Nuclear Regulatory Commission (NRC) introduced technical guidance for allowing burnup credit in the criticality safety analysis of casks for transporting or storing spent fuel from pressurized water reactors. This paper presents the recommendations embodied by the current NRC guidance, discusses associated technical issues, and reviews information needs and industry priorities for expanding the scope and content of the guidance. Allowable analysis approaches for burnup credit must account for the fuel irradiation variables that affect spent fuel reactivity, including the axial and horizontal variation of burnup within fuel assemblies. Consistent with international transport regulations, the burnup of each fuel assembly must be verified by pre-loading measurements. The current guidance limits the credited burnup to no more than 40 GWd/MTU and the credited cooling time to five years, imposes a burnup offset for fuels with initial enrichments between 4 and 5 wt% 235U, does not include credit for fission products, and excludes burnup credit for damaged fuels and fuels that have used burnable absorbers. Burnup credit outside these limits may be considered when adequately supported by technical information beyond that reviewed to-date by the NRC staff. The guidance further recommends that residual subcritical margins from the neglect of fission products, and any other nuclides not credited in the licensing-basis analysis, be estimated for each cask design and compared against estimates of the maximum reactivity effects associated with remaining computational uncertainties and potentially nonconservative modeling assumptions. The NRC's Office of Nuclear Regulatory Research is conducting a research program to help develop the technical information needed for refining and expanding the evolving guidance. Cask vendors have announced plans to submit the first NRC license applications for burnup credit later this year

  13. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Lake, W.H.

    1989-01-01

    The traditional assumption used in evaluating criticality safety of spent fuel cask is that the spent fuel is as reactive as when it was fresh (new). This is known as the fresh fuel assumption. It avoids a number of calculational and verification difficulties, but could take a heavy toll in decreased efficiency. The alternative to the fresh fuel assumption is called burnup credit. That is, the reduced reactivity of spent fuel that comes about from depletion of fissile radionuclides and net increase in neutron absorbers (poisons) is taken into account. It is recognizable that the use of burnup credit will in fact increase the percentage of unacceptable or non-specification fuel available for misloading. This could reduce individual cask safety margins if current practices with respect to loading procedures are maintained. As such, additional operational, design, analysis, and validation requirements should be established that, as a minimum, compensate for any potential reduction in fuel loading safety margin. This method is based on a probabilistic (PRA) approach and is called a relative risk comparison. The method assumes a linear risk model, and uses a selected probability function to compare the system of interest and an acceptable reference system by varying the features of each to assess effects on system safety. While risk is the product of an event probability and its consequence, the consequences of criticality in a cask are considered to be both unacceptable and the same, regardless of the initiating sequence. Therefore, only the probability of the event is considered in a relative risk evaluation

  14. Research on Spent Fuel Storage and Transportation in CRIEPI (Part 2 Concrete Cask Storage)

    Energy Technology Data Exchange (ETDEWEB)

    Koji Shirai; Jyunichi Tani; Taku Arai; Masumi Watatu; Hirofumi Takeda; Toshiari Saegusa; Philip L. Winston

    2008-10-01

    Concrete cask storage has been implemented in the world. At a later stage of storage period, the containment of the canister may deteriorate due to stress corrosion cracking phenomena in a salty air environment. High resistant stainless steels against SCC have been tested as compared with normal stainless steel. Taking account of the limited time-length of environment with certain level of humidity and temperature range, the high resistant stainless steels will survive from SCC damage. In addition, the adhesion of salt from salty environment on the canister surface will be further limited with respect to the canister temperature and angle of the canister surface against the salty air flow in the concrete cask. Optional countermeasure against SCC with respect to salty air environment has been studied. Devices consisting of various water trays to trap salty particles from the salty air were designed to be attached at the air inlet for natural cooling of the cask storage building. Efficiency for trapping salty particles was evaluated. Inspection of canister surface was carried out using an optical camera inserted from the air outlet through the annulus of a concrete cask that has stored real spent fuel for more than 15 years. The camera image revealed no gross degradation on the surface of the canister. Seismic response of a full-scale concrete cask with simulated spent fuel assemblies has been demonstrated. The cask did not tip over, but laterally moved by the earthquake motion. Stress generated on the surface of the spent fuel assemblies during the earthquake motion were within the elastic region.

  15. Considerations in the transportation of spent fuel and other radioactive materials by sea

    International Nuclear Information System (INIS)

    Rhoads, R.E.; Heaberlin, S.W.; Brobst, W.A.

    1978-05-01

    A study was made of what would happen to casks and plutonium packages in the unlikely event of a serious sea accident involving shipments of spent fuel or plutonium, and what the radiological consequences might be to man through the marine food chain. Four postulated loss scenarios were considered: loss on the continental shelf and in the deep ocean of an undamaged package, and a package exposed to a severe fire. Conservative estimates of the maximum population doses are comparable to those expected from natural background sources. Areas where futher research is needed are identified. 3 tables

  16. Spent fuel transportation on highways: the radioactive dose to the traffic

    International Nuclear Information System (INIS)

    Yadigaroglu, G.

    1975-01-01

    The radioactive exposure of the traffic moving on the same highway as spent fuel shipments has been neglected in the past. Methods developed for calculating peak exposures, the number of individuals receiving a dose in excess of a certain limiting value, and the cumulative population doses for the occupants of the vehicles under a variety of highway and accident conditions allow comparisons to the corresponding stationary-population doses. Consideration of both routine direct-radiation exposures and accidental releases indicates that the traffic doses can be of equal or greater importance than the stationary-population doses

  17. Assessment of irradiated nuclear spent fuel transport in the Commonwealth of independent states

    International Nuclear Information System (INIS)

    Raisonnier, D.; Lemogne, A.; Semin-Vadov, S.; Koshik, Y.

    1998-01-01

    Irradiated nuclear fuels (INFs) are transported regularly in and between the New Independent States (NIS). Under the TACIS Programme (Technical Assistance to the Commonwealth of Independent States), the European Commission has funded this project devoted to the assessment of transports of INF in Russia, Kazakhstan and Ukraine in collaboration with local State Organizations, competent Authorities, nuclear facilities and transport companies. This paper will present the development of this project during the two years of its implementation as well as its main results. (authors)

  18. Current status and future considerations for a transportation system for spent fuel and radioactive waste

    International Nuclear Information System (INIS)

    Anderson, R.T.; Darr, D.G.; Godfrey, W.L.; Keely, R.B.; Lusk, E.C.; Peterson, R.W.; Ridihalgh, J.L.; Shallo, F.A.; Young, M.

    1978-02-01

    This report is part of the OWI Transportation/Logistics systems analysis of problems associated with shipping these wastes to waste terminal storage facilities. It covers governmental regulations and functional responsibilities, highway and rail transportation status and economic considerations, assessment of present industry capabilities and business-related considerations, important receiving facility considerations, necessary engineering and licensing-related aspects of packaging systems, and essential elements of reprocessing plant waste activities including packaging and transportation

  19. Workshop on spent fuel performance, radionuclide chemistry and geosphere transport parameters, Lidingoe 2008: Overview and evaluation of recent SKB procedures

    International Nuclear Information System (INIS)

    Meinrath, Guenther; Stenhouse, Mike; Brown, Paul; Ekberg, Christian; Jegou, Christophe; Nitsche, Heino

    2009-08-01

    The safety assessment for disposal of spent nuclear fuel canister in the Swedish bedrock should thoroughly address the time period after a containment failure. Such a failure could be expected as a result of corrosion damage or mechanical failure due to rock movement. This report mainly covers some issues connected to parameters used for radionuclide transport calculations in the areas of spent fuel performance (for fuel in contact with groundwater), radionuclide chemistry, and sorption and geosphere transport parameters. Some examples of topics that are elaborated in some detail include statistical treatment of measurement data (for sorption measurements), handling of uncertainties in speciation calculations, use of triangular distributions in safety assessment and physical processes in connection with spent fuel aging. The results emerged from discussions among international experts at a workshop in May 2008. The purpose of this work is providing an overview of ongoing work within the Swedish Nuclear Fuel and Waste Management Co. (SKB), to provide ideas and suggestions for methodology development and to develop review capability within the SSM. The authors conclude that SKB's treatment of uncertainty in speciation calculations has improved, but that additional efforts in the area of error propagation are recommended. In efforts to condense the scope of utilised thermodynamic databases, the authors recommend that exclusion criteria should be explicitly stated. In the area of sorption, there is a need for more thorough analysis of errors in order to establish uncertainty ranges. The most essential improvements concern dose-limiting nuclides (e.g. Ra-226). Triangular distributions are often featured in SKB safety assessment, but it is not clear that the use of such distributions is based on a firm understanding of its properties. Regarding fuel performance, while safety assessment parameters are supported by measurement data there is still a need for better

  20. Workshop on spent fuel performance, radionuclide chemistry and geosphere transport parameters, Lidingoe 2008: Overview and evaluation of recent SKB procedures

    Energy Technology Data Exchange (ETDEWEB)

    Meinrath, Guenther; Stenhouse, Mike; Brown, Paul; Ekberg, Christian; Jegou, Christophe; Nitsche, Heino

    2009-08-15

    The safety assessment for disposal of spent nuclear fuel canister in the Swedish bedrock should thoroughly address the time period after a containment failure. Such a failure could be expected as a result of corrosion damage or mechanical failure due to rock movement. This report mainly covers some issues connected to parameters used for radionuclide transport calculations in the areas of spent fuel performance (for fuel in contact with groundwater), radionuclide chemistry, and sorption and geosphere transport parameters. Some examples of topics that are elaborated in some detail include statistical treatment of measurement data (for sorption measurements), handling of uncertainties in speciation calculations, use of triangular distributions in safety assessment and physical processes in connection with spent fuel aging. The results emerged from discussions among international experts at a workshop in May 2008. The purpose of this work is providing an overview of ongoing work within the Swedish Nuclear Fuel and Waste Management Co. (SKB), to provide ideas and suggestions for methodology development and to develop review capability within the SSM. The authors conclude that SKB's treatment of uncertainty in speciation calculations has improved, but that additional efforts in the area of error propagation are recommended. In efforts to condense the scope of utilised thermodynamic databases, the authors recommend that exclusion criteria should be explicitly stated. In the area of sorption, there is a need for more thorough analysis of errors in order to establish uncertainty ranges. The most essential improvements concern dose-limiting nuclides (e.g. Ra-226). Triangular distributions are often featured in SKB safety assessment, but it is not clear that the use of such distributions is based on a firm understanding of its properties. Regarding fuel performance, while safety assessment parameters are supported by measurement data there is still a need for better

  1. Impacts of transportation regulations on spent fuel and high level waste cask design

    International Nuclear Information System (INIS)

    Lake, W.H.

    1992-01-01

    The regulation of radioactive material transport has a long and successful history. Prior to 1966, these activities were regulated by the Interstate Commerce Commission (ICC) Bureau of Explosives (the ICC was predecessor to the Department of Transportation (DOT)). In 1966, the Atomic Energy Commission (AEC) developed what is now 10 CFR 71, concurrently with the development of similar international standards. In 1975, the AEC was reorganized and the Nuclear Regulatory Commission (NRC) was established as an independent regulatory commission. The NRC was given responsibility for the regulation of commercial use of radioactive materials, including transportation. This paper discusses various aspects of the NRC's role in the transport of radioactive material as well as its role in the design and certification of casks necessary to the transport of this material

  2. Selective removal of carbon-14 from ion exchange resins using supercritical carbon dioxide

    International Nuclear Information System (INIS)

    Dias, S.A.; Krasznai, J.P.

    1996-01-01

    Ion exchange resins (IX) are used extensively in CANDU-PHWR (Canada Deuterium Uranium - Pressurized Heavy Water Reactor) and other reactor systems worldwide to remove ionic contaminants from various coolant circuits. Spent IX resins represent a significant volume of low and intermediate level radioactive waste. The presence of long-lived C-14 which is found in significant quantities in IX resins from CANDU reactors, complicates the disposal of these resins. Several experiments were conducted with carbon dioxide under subcritical and supercritical conditions to determine the feasibility of removing C-14 present as carbonate and/or bicarbonate on IX resins. It has been established that resins containing inorganic C-14 undergo rapid isotopic exchange when exposed to inactive carbon dioxide under supercritical conditions. This treatment reduces the C-14 to the limits of detection and leaves other radioisotopes on the resins largely unaffected. This selective and highly efficient means to remove long-lived C-14 activity from CANDU spent IX resins allows the resin waste to be reclassified as low level waste. This lower classification simplifies the handling, transportation and eventual disposal of IX resins which translates to a very significant cost saving. Since the process is selective the C-14 can be enriched and recovered for commercial purposes

  3. Operational considerations in specifying legal weight vehicles for the highway transport of spent nuclear fuel

    International Nuclear Information System (INIS)

    Hill, C.V.; Rutenkroger, E.O.; Ratledge, J.E.

    1990-01-01

    This paper presents the results of a research project in which tractor manufacturers and carrier companies were interviewed to gather information on operational concerns in specifying a tractor to haul legal weight spent fuel casks. The system was assumed to operate very close to the 80,000 pound legal weight limit. Safety, performance, reliability, and maintainability of equipment were factors given particular attention. The interaction between driver fatigue, safety, and equipment was also discussed. Innovative operating strategies that could save weight were discussed. The paper concluded that operational considerations require that planners working with standard off-the-shelf tractor equipment should allow at least 17,350 pounds for the weight of the tractor as a starting point from which further weight reduction analysis can proceed. 4 refs., 3 figs

  4. Technology Development for Integrated Safety Test of Spent Nuclear Fuel Transportation and Storage System

    International Nuclear Information System (INIS)

    Seo, Kiseog; Seo, J. S.; Lee, J. C.

    2012-05-01

    A dedicated review on the U. S. NRC Regulation 10 CFR Part 72 and regulatory guide NUREG/1536 has been performed. The safety requirements for spent nuclear fuel dry storage cask are analyzed and summarized in structural, thermal, shielding, criticality, materials, tests and maintenance aspects. Also a guideline for preparing the safety analysis report is provided. The heat flow analysis was performed by varying the dimensions of the heat flow test facility. From the heat flow analysis for the test facility, as the test facility became test facility. From the heat flow analysis for the test facility, as the test facility became bigger; the thermal effect became smaller. Therefore, the dimensions of the heat flow test facility was designed with 5m Χ 5m Χ 6m(H). Analyses of heat transfer characteristics and mechanism for spent PWR fuel assemblies, option study for production of the effective thermal conductivity and option study for effective thermal conductivity test have been performed to obtain the basic data for production of the effective thermal conductivity. It became clear that the diffusion coefficient of chloride ion of concrete remarkably increases along with the temperature rise, and that there is a linear relation between the logarithm values of the diffusion coefficients and the reciprocal of the temperature. It is understood to be able to express the temperature dependency of the diffusion coefficient roughly by an Arrhenius equation as the velocity coefficient is provided as the diffusion coefficient. The specifications and characteristics of storage facilities under operation including dual purpose casks were investigated. Components subject to material degradation were examined. Based on literature survey, investigating a drop analysis incorporating with material degradation, the basic data to develop an analysis methodology was obtained

  5. Development of a Computer Program for an Analysis of the Logistics and Transportation Costs of the PWR Spent Fuels in Korea

    International Nuclear Information System (INIS)

    Cha, Jeong Hun; Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won

    2009-01-01

    It is expected that a substantial amount of spent fuels will be transported from the four nuclear power plant (NPP) sites in Korea to a hypothetical centralized interim storage facility or a final repository in the near future. The cost for the transportation is proportional to the amount of spent fuels. In this paper, a cost estimation program is developed based on the conceptual design of a transportation system and a logistics analysis. Using the developed computer program, named as CASK, the minimum capacity of a centralized interim storage facility (CISF) and the transportation cost for PWR spent fuels are calculated. The PWR spent fuels are transported from 4 NPP sites to a final repository (FR) via the CISF. Since NPP sites and the CISF are located along the coast, a sea-transportation is considered and a road-transportation is considered between the CISF and the FR. The result shows that the minimum capacity of the interim storage facility is 15,000 MTU

  6. A Transportation Risk Assessment Tool for Analyzing the Transport of Spent Nuclear Fuel and High-Level Radioactive Waste to the Proposed Yucca Mountain Repository

    International Nuclear Information System (INIS)

    Best, Ralph; Winnard, T.; Ross, S.; Best, R.

    2001-01-01

    The Yucca Mountain Transportation Database was developed as a data management tool for assembling and integrating data from multiple sources to compile the potential transportation impacts presented in the Draft Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada (DEIS). The database uses the results from existing models and codes such as RADTRAN, RISKIND, INTERLINE, and HIGHWAY to estimate transportation-related impacts of transporting spent nuclear fuel and high-level radioactive waste from commercial reactors and U. S. Department of Energy (DOE) facilities to Yucca Mountain. The source tables in the database are compendiums of information from many diverse sources including: radionuclide quantities for each waste type; route and route characteristics for rail, legal-weight truck, heavy haul. truck, and barge transport options; state-specific accident and fatality rates for routes selected for analysis; packaging and shipment data by waste type; unit risk factors; the complex behavior of the packaged waste forms in severe transport accidents; and the effects of exposure to radiation or the isotopic specific effects of radionclides should they be released in severe transportation accidents. The database works together with the codes RADTRAN (Neuhauser, et al, 1994) and RISKlND (Yuan, et al, 1995) to calculate incident-free dose and accident risk. For the incident-free transportation scenario, the database uses RADTRAN and RISKIND-generated data to calculate doses to offlink populations, onlink populations, people at stops, crews, inspectors, workers at intermodal transfer stations, guards at overnight stops, and escorts, as well as non-radioactive pollution health effects. For accident scenarios, the database uses RADTRAN-generated data to calculate dose risks based on ingestion, inhalation, resuspension, immersion (cloudshine), and groundshine as

  7. Transportation of Spent Nuclear Fuel and High Level Waste to Yucca Mountain: The Next Step in Nevada

    International Nuclear Information System (INIS)

    Sweeney, Robin L.; Lechel, David J.

    2003-01-01

    In the U.S. Department of Energy's ''Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada,'' the Department states that certain broad transportation-related decisions can be made. These include the choice of a mode of transportation nationally (mostly legal-weight truck or mostly rail) and in Nevada (mostly rail, mostly legal-weight truck, or mostly heavy-haul truck with use of an associated intermodal transfer station), as well as the choice among alternative rail corridors or heavy-haul truck routes with use of an associated intermodal transfer station in Nevada. Although a rail line does not service the Yucca Mountain site, the Department has identified mostly rail as its preferred mode of transportation, both nationally and in the State of Nevada. If mostly rail is selected for Nevada, the Department would then identify a preference for one of the rail corridors in consultation with affected stakeholders, particularly the State of Nevada. DOE would then select the rail corridor and initiate a process to select a specific rail alignment within the corridor for the construction of a rail line. Five proposed rail corridors were analyzed in the Final Environmental Impact Statement. The assessment considered the impacts of constructing a branch rail line in the five 400-meter (0.25mile) wide corridors. Each corridor connects the Yucca Mountain site with an existing mainline railroad in Nevada

  8. One approach to accepting and transporting spent fuel from early-generation reactors with short fuel assemblies

    International Nuclear Information System (INIS)

    Peterson, R.W.; Bentz, E.J. Jr.; Bentz, C.B.

    1993-01-01

    In the early days of development of commercial nuclear power reactors in the U.S., the overall length and uranium loading of the fuel assemblies were considerably less than those of later generation facilities. In turn, some of these early facilities were designed for handling shorter casks than currently-certified casks. The spent fuel assemblies from these facilities are nearly all standard fuel within the definition in the Standard Contract (10 CFR 961) between the utilities and the U.S. Department of Energy (DOE) (the Big Rock Point fuel cross-section is outside the standard fuel dimension), and the utilities involved hold early delivery rights under DOE's oldest-fuel-first (OFF) allocation scenario. However, development of casks suitable for satisfying the acceptance and transportation requirements of some of these facilities is not currently underway in the DOE Cask System Development Program (CSDP). While the total MTU of these fuels is relatively small compared to the total program, the number of assemblies to be transported is significant, especially in the early years of operation according to the OFF allocation scenario. We therefore perceive a need for DOE to develop an approach and to implement plans to satisfy the unique acceptance and transportation requirements of these facilities. One such approach is outlined below. (author)

  9. Analysis of alternative transportation methods for radioactive materials shipments including the use of special trains for spent fuel and wastes

    International Nuclear Information System (INIS)

    Smith, D.R.; Luna, R.E.; Taylor, J.M.

    1978-01-01

    Two studies were completed which evaluate the environmental impact of radioactive material transport. The first was a generic study which evaluated all radioactive materials and all transportation modes; the second addressed spent fuel and fuel-cycle wastes shipped by truck, rail and barge. A portion of each of those studies dealing with the change in impact resulting from alternative shipping methods is presented in this paper. Alternatives evaluated in each study were mode shifts, operational constraints, and, in generic case, changes in material properties and package capabilities. Data for the analyses were obtained from a shipper survey and from projections of shipments that would occur in an equilibrium fuel cycle supporting one hundred 1000-MW(e) reactors. Population exposures were deduced from point source radiation formulae using separation distances derived for scenarios appropriate to each shipping mode and to each exposed population group. Fourteen alternatives were investigated for the generic impact case. All showed relatively minor changes in the overall radiological impact. Since the radioactive material transport is estimated to be fewer than 3 latent cancer fatalities (LCF) for each shipment year (compared to some 300,000 yearly cancer fatalities or 5000 LCF's calculated for background radiation using the same radiological effects model), a 15% decrease caused by shifting from passenger air to cargo air is a relatively small effect. Eleven alternatives were considered for the fuel cycle/special train study, but only one produced a reduction in total special train baseline LCF's (.047) that was larger than 5%

  10. Facility - transportation system interface at COGEMA (experience in taking delivery of spent fuel)

    International Nuclear Information System (INIS)

    Lenail, B.

    1983-01-01

    COGEMA has long experience in taking delivery of spent fuel from LWR's. This activity is now fully industrialized. Great care is taken by COGEMA and their affiliates in this business, generally beyond the normal regulatory limits, and this is quite normal, COGEMA wanting to prevent, by all possible means, causes of incidents or accidents which could destroy the position it has established and the confidence placed in it by many customers. It is clear that the situation is quite different at reactors and at reprocessing plants; the limited number of flasks handled at reactors (in general five per year) is generally not enough to lead reactor operators to invest in sophisticated equipment and to adopt procedures allowing extremely small exposures of their personnel, while the great number of flasks received each year in a plant like La Hague (not fewer than 200 per year) justifies the investment in more specialized equipment (automation, remote operation, more sophisticated tools, etc.) and allows the personnel to be very well trained to that particular type of work. These are factors leading to very limited exposures of the personnel although the work performed in a repocessing plant is generally more complex (deeper decontamination, closer work, maintenance, etc.) than in a reactor

  11. Regional cooperation for transport of nuclear and radioactive material. Safety, security and safeguards related regulatory tasks concerning international spent fuel shipments

    International Nuclear Information System (INIS)

    Koblinger, László; Horváth, Kristóf

    2016-01-01

    The safety, security and safeguards related regulatory tasks of international spent fuel transports are to be performed according to international conventions; additionally, Hungary as a member state of the European Union shall comply with the relevant EURATOM requirements. The international requirements are reflected in the national legislations of the countries concerned by the transports of research reactor spent fuel under the Global Threat Reduction Initiative repatriation project. The successful realization of these shipments demanded long-lasting preparation with thorough cooperation and conduct with clockwork precision by the regulatory bodies within states and regionally

  12. German physical protection concept for the storage of spent fuel elements in transport and storage casks

    International Nuclear Information System (INIS)

    Weil, L.; Maier, R.

    2005-01-01

    Full text: In Germany, the legal regulations and requirements derived from rules and guidelines for the protection of storage facilities for spent fuel elements from disruptive action or other inference by third parties are structured hierarchically. The Atomic Energy Act constitutes the top level. It is supported by federal ordinances. The next level down is formed by the rules and guidelines. The storage of nuclear fuels may only be authorized, according to the provisions of the Atomic Energy Act, if the required protection from disruptive action or other interference by third parties can be guaranteed as following: it must be possible to prevent any danger to life and health due to a substantial amount of direct radiation or due to the release of a substantial amount of radioactive material; it must be possible to prevent singular or repeated acts of stealing nuclear fuels in such amounts that a critical accumulation can be produced directly without reprocessing and enrichment. Knowing that nuclear installations cannot be protected from every possible interference, physical protection is focused on basic security standards, the so-called design basic threat (DBT), departing from the assumed interference. DBT is regularly reviewed by the competent federal authorities and authorities of the states and are revised on the basis of newly gained knowledge, if necessary, such as in the wake of the terrorist attacks in the U.S. on September 11, 2001. The operator must guarantee and give proof of a sufficient level of physical protection of the plant. The sole physical protection measures implemented by the operator cannot ensure the required protection from other interference by third parties for an unlimited time span. The concept therefore requires additional physical protection measures by the police. (author)

  13. Interfaces between transport and geological disposal systems for high level radioactive waste and spent nuclear fuel

    International Nuclear Information System (INIS)

    1994-09-01

    This document is an IAEA publication which identifies and discusses the interfaces and the interface requirements between high level waste, the waste transport system used for carriage of the waste to the disposal facility, and the high level waste disposal facility. The development of this document was prompted in part by the initiatives in various Member States to select, characterize and design the facilities for potential high level waste geological repositories. These initiatives have progressed to the point where an international document would be useful in calling attention to the need for establishing, in a systematic way, interfaces and interface requirements between the transport systems to be used and the waste disposal packages and geological repository. Refs, figs and tabs

  14. Guidebook on spent fuel storage

    International Nuclear Information System (INIS)

    1984-01-01

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

  15. Small gas leakage rate measuring and monitoring system for spent fuel transport/storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Asano, Ryoji [Hitachi Zosen Corp., Tokyo (Japan); Aritomi, Masanori; Matsuzaki, Mitsuo; Kawa, Tsunemichi

    1997-12-01

    A containment function of transport and/or storage casks of radioactive materials is essential to prevent the materials from being released excessively into the environment. It is not practical for containment tests to measure directly the radioactivity release so that gas volumetric leakage rates are usually assessed and gas pressure decrease or increase method is usually applied. As gas flow model for evaluation, the ISO standards has deleted the concept of choked flow which is adopted by ANSI N14.5. Provided that the choked flow is not adopted to the leakage rate evaluation, the criteria of the test should be severe, and a new leakage rate measuring system with high accuracy and reasonable measuring time is required. Transport casks are often inspected in a temporary cask-storage facility where simultaneous measurement of many casks is required. In a storage cask system, multiple casks are monitored on their containment function during a storage period, and the method for simultaneous monitoring at many points for long term is required. In this study, two kinds of small gas leakage rate measuring systems are developed. One is to measure gas leakage rates directly and is called `flow measuring system`, which can measure gas leakage rate of 10{sup -4} to 10{sup -2} cm{sup 3}/s with high accuracy and short measuring time. The other is to measure the pressure decreasing rate and is called `pressure decreasing rate measuring system`, which can monitor the pressure change at many points simultaneously. (author)

  16. Estimation of Shielding Thickness for a Prototype Department of Energy National Spent Nuclear Fuel Program Transport Cask

    Energy Technology Data Exchange (ETDEWEB)

    SANCHEZ,LAWRENCE C.; MCCONNELL,PAUL E.

    2000-07-01

    Preliminary shielding calculations were performed for a prototype National Spent Nuclear Fuel Program (NSNFP) transport cask. This analysis is intended for use in the selection of cask shield material type and preliminary estimate of shielding thickness. The radiation source term was modeled as cobalt-60 with radiation exposure strength of 100,000 R/hr. Cobalt-60 was chosen as a surrogate source because it simultaneous emits two high-energy gammas, 1.17 MeV and 1.33 MeV. This gamma spectrum is considered to be large enough that it will upper bound the spectra of all the various spent nuclear fuels types currently expected to be shipped within the prototype cask. Point-kernel shielding calculations were performed for a wide range of shielding thickness of lead and depleted uranium material. The computational results were compared to three shielding limits: 200 mrem/hr dose rate limit at the cask surface, 50 mR/hr exposure rate limit at one meter from the cask surface, and 10 mrem/hr limit dose rate at two meters from the cask surface. The results obtained in this study indicated that a shielding thickness of 13 cm is required for depleted uranium and 21 cm for lead in order to satisfy all three shielding requirements without taking credit for stainless steel liners. The system analysis also indicated that required shielding thicknesses are strongly dependent upon the gamma energy spectrum from the radiation source term. This later finding means that shielding material thickness, and hence cask weight, can be significantly reduced if the radiation source term can be shown to have a softer, lower energy, gamma energy spectrum than that due to cobalt-60.

  17. Private shippers' efforts to transport spent nuclear fuel and waste by rail: the Interstate Commerce Commission cases

    International Nuclear Information System (INIS)

    Brown, O.F. II

    1978-01-01

    While the nuclear industry is beset on many sides by opposition and uncertainty, notable success has been achieved to date by fighting railroads that would impose onerous restrictions on the transportation of nuclear materials on their own battleground. Regardless of how commercial spent nuclear fuel is ultimately handled, it must eventually be shipped from reactor sites. The most efficient method for doing this is generally by railroad. The railroads of the Nation have contended either that they are not common carriers for the transportation of radioactive materials, or if common carriers, that the carriage must be by the use of special trains and special operating procedures at great additional expense. The rationale for this railroad position has been principally one of concern with regard to the hazards presented and the adequacy of safety regulations and insurance-indemnity coverage. The nuclear industry, realizing that, if the railroads prevailed, the added cost per year by the mid-1980s could be about $600 million and that extensive safety precautions had been provided, challenged the railroads' positions before the Interstate Commerce Commission. This paper will describe the position of the nuclear industry in opposing unnecessary restrictions sought to be imposed by the railroads, including the uncertainties presented by the lack of published tariffs, the risks of carriage with and without added requirements, the insurance-indemnity coverage available, and the need for railroad transportation. As of this time, the Interstate Commerce Commission has concurred with the position of the nuclear shippers and has found the railroads' positions unreasonable and in violation of applicable laws. This legal battle, however, will continue until finally concluded by the ICC or by the courts if appeals are taken

  18. Properties of the Carboxylate ion exchange resins

    International Nuclear Information System (INIS)

    Allard, Bert; Dario, Maarten; Boren, Hans; Torstenfelt, Boerje; Puigdomenech, Ignasi; Johansson, Claes

    2002-09-01

    Weakly acidic, carboxylic resin has been selected, together with strong base anion resins, for water purification at the Forsmark 1 and 2 reactors. For the strong (but not the weak) ion exchange resin the Nuclear Power Inspectorate has given permission to dispose the spent resins in the SFR 1 (the Final Repository for Radioactive Operational Waste). This report gives a review of the carboxylic resins and comes to the conclusion that the resins are very stable and that there should not exist any risks for increased leaching of radionuclides from SFR 1 if these resins are disposed (compared to the strong resins)

  19. Metal cask RT-5000 for the dry storage and transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vorobyov, A.I.; Kazeev, V.G.; Krayev, V.S.; Shcherbina, A.N.; Churikov, Y.I. [All-Russian Research Inst. of Technical Physics, Snezhinsk (Russian Federation)

    2003-05-01

    Presentation of new-type cask, developed at RFNC-VNIITF, is in the article. The prototype model of the shipping cask was subjected to tests imitating normal shipment conditions (free fall, pressing, and impact) and to tests imitating emergency situation during shipment (a drop from the 9-m height onto a pin is replaced by acceleration of the shipping cask at a guide rail of the rocket-catapult installation (RCI), a 1-m drop onto a pin, heat tests a 30-minutes fire at the temperature of for 8500 C, submergence to the depth of 15 and 200 meters). After each test the hermeticity preservation is examined. Parallel with the real testing, a mathematical simulation of physical processes induced by the corresponding tests was conducted at the RFNC-VNIITF. The required parameters obtained from the tests are used to calibrate the calculation methods. As a result it has been possible to obtain a good agreement between the results of calculations and experiments; this will allow the mathematic simulation to be used wider. The advantage of the RT-5000 metal cask in comparison with metal-concrete analogs are as follows: SFA are placed into the RT-5000 entirely without cutting into two bunches of fuel elements; the expensive hot doom is not required for automatic cutting the SFA and for loading the bunches of fuel elements into intermediate cases; the possibility remains to transport the RT-5000 without reloading SFA after 50-year storage, although this is a problem for the metal-concrete casks.

  20. Comparative assessment of spent nuclear fuel transportation routes using risk factors and a geographic information system

    International Nuclear Information System (INIS)

    Toth, D.M.

    1996-01-01

    The assessment of potential alternative routes was simplified through the use of six comparative risk factors evaluated using detailed, route-specific data. The route and environmental attributes varied strongly with location and were developed from national, state, and local sources. The six comparative factors were risk measures of incident-free transportation radiation exposure, radiological accident population exposure, nonradiological accidents, contamination of environmental sensitive areas, environmental justice for minority populations, and environmental justice for low-income populations. An assessment of four real North-Central Florida routes provided a sample implementation of the analysis tools and risk factors. The assessment routes, consisting of common beginning and end locations, included an interstate highway, a rural highway, a mostly urban highway, and a combination interstate highway with rural bypass. This route comparative assessment study predicted that the interstate highway, despite a higher population density, greater traffic volume, and greater number of vehicular fatality accidents, would present the lowest cumulative risk. On the contrary, the rural highway route, characterized as having the lowest population density, minimal vehicle traffic volume, and the lowest percentages of minority and low-income populations, displayed the highest cumulative risk measure. Factors contributing to the high risk for the rural highway route included greater route length, higher vehicular fatality accident rates per vehicle mile traveled, and the close proximity to environmentally sensitive areas. This route comparative assessment study predicted that the interstate highway, despite a higher population density, greater traffic volume, and greater number of vehicular fatality accidents, would present the lowest cumulative risk. On the contrary, the rural highway route, characterized as having the lowest population density, minimal vehicle traffic volume

  1. Determination of 63Ni and 59Ni in spent ion-exchange resin and activated charcoal from the IEA-R1 nuclear research reactor.

    Science.gov (United States)

    Taddei, M H T; Macacini, J F; Vicente, R; Marumo, J T; Sakata, S K; Terremoto, L A A

    2013-07-01

    A radiochemical method has been adapted to determine (59)Ni and (63)Ni in samples of radioactive wastes from the water cleanup system of the IEA-R1 nuclear research reactor. The process includes extraction chromatographic resin with dimethylglyoxime (DMG) as a functional group. Activity concentrations of (59)Ni and (63)Ni were measured, respectively, by X-ray spectrometry and liquid scintillation counting, whereas the chemical yield was determined by ICP-OES. The average ratio of measured activity concentrations of (63)Ni and (59)Ni agree well with theory. Copyright © 2013 Elsevier Ltd. All rights reserved.

  2. An investigation into the hazards associated with the maritime transport of spent nuclear reactor fuel to the British Isles

    International Nuclear Information System (INIS)

    1980-01-01

    Interim results are presented from an investigation into the potential hazard from maritime transport of spent reactor fuel. From a review of official safety studies the most severe accident is identified as a prolonged shipboard fire of 9 hours or more. According to studies performed for the International Atomic Energy Agency by the Batelle Laboratories such a fire could fail all fuel elements and release volatile radionuclides such as caesium to the environment. The consequences of such an accident are investigated for a release to the Irish Sea from a fire damaged vessel. Consequences are analysed for a release to the continental shelf waters following sinking, and also for an atmospheric release close to a conurbation. The port of Barrow is taken as an example. The report concludes that either of these events could have catastrophic consequences: the Irish Sea might have to be closed to fisheries and in the case of an atmospheric release large scale evacuation would be necessary to prevent loss of life. (author)

  3. Trial evaluation on criticality safety of the fuel assemblies at falling accident as spent fuel transport and storage cask

    International Nuclear Information System (INIS)

    Tadano, Tomoaki

    2016-01-01

    The authors conducted critical safety assessment on the supposed event at the time of a fall accident of cask, and examined the influence on criticality safety. If the spacer of fuel assembly is sound, it is assumed that the pitch of fuel rod interval changes, and if the spacer is broken, it is assumed that the fuel rod is unevenly distributed in the basket. For the critical calculation of fuel assembly basket system, they performed it using a calculation code. For both of the single cell and assembly, calculation results showed an increase in the effective multiplication factor of reactivity of 2-3%. When this reactivity is applied to the criticality analysis result of PWR fuel assembly, the value approaches to the limit 0.95 of the neutron effective multiplication factor keff. However, the keff when new fuel is loaded is sufficiently lower than 0.93. Therefore, it is unlikely that the criticality analysis result approaches to 0.95 at all burnups, and the possibility to become criticality is very low in actual spent fuel transport. When considering the reactivity of this research, it is possible that the design condition for the assumption of novel fuel loading becomes severer. Furthermore, criticality analysis under non-uniform pitch will become necessary, and criticality safety analysis for BWR fuel with heterogeneous enrichment degree and burnup degree will become also necessary. (A.O.)

  4. Spent fuel transport and storage system for NOK: The TN52L, TN97L, TN24 BHL and TN24 GB casks

    International Nuclear Information System (INIS)

    Wattez, L.; Verdier, A.; Monsigny, P.-A.

    2007-01-01

    NOK nuclear power plants in Switzerland, LEIBSTADT (KKL) BWR nuclear power plant and BEZNAU (KKB) PWR nuclear power plant have opted to ship spent fuel to a central facility called ZWILAG for interim storage. In the mid-nineties, COGEMA LOGISTICS was contracted by KKL for the supply of the TN52L and TN97L transport and storage casks for BWR fuel types. In 2003, KKL also ordered from COGEMA LOGISTICS the supply of six TN24 BHL transport and storage casks. This paper shows how all the three cask designs have responded to the KKL needs to ship and store BWR spent fuel. In addition, it highlights the already significant operational feedback of the TN52L and TN97L casks by the KKL and ZWILAG operators. In 2004, NOK also ordered three TN24 GB transport and storage casks for PWR fuel types. These casks are presently being manufactured. (author)

  5. Methods for estimating costs of transporting spent fuel and defense high-level radioactive waste for the civilian radioactive waste management program

    International Nuclear Information System (INIS)

    Darrough, M.E.; Lilly, M.J.

    1989-01-01

    The US Department of Energy (DOE), through the Office of Civilian Radioactive Waste Management, is planning and developing a transportation program for the shipment of spent fuel and defense high-level waste from current storage locations to the site of the mined geologic repository. In addition to its responsibility for providing a safe transportation system, the DOE will assure that the transportation program will function with the other system components to create an integrated waste management system. In meeting these objectives, the DOE will use private industry to the maximum extent practicable and in a manner that is cost effective. This paper discusses various methodologies used for estimating costs for the national radioactive waste transportation system. Estimating these transportation costs is a complex effort, as the high-level radioactive waste transportation system, itself, will be complex. Spent fuel and high-level waste will be transported from more than 100 nuclear power plants and defense sites across the continental US, using multiple transport modes (truck, rail, and barge/rail) and varying sizes and types of casks. Advance notification to corridor states will be given and scheduling will need to be coordinated with utilities, carriers, state and local officials, and the DOE waste acceptance facilities. Additionally, the waste forms will vary in terms of reactor type, size, weight, age, radioactivity, and temperature

  6. In vitro comparison rate of dental root canal transportation using two single file systems on the simulated resin blocks

    Directory of Open Access Journals (Sweden)

    Mohammad Javad Etesami

    2016-07-01

    Full Text Available Background and Aims: Cleaning and shaping is one of the most important stages in endodontic treatment. Single-file systems save time and reduce the risk of transmission of pathogens. This in vitro study was aimed to compare the rate of canal transportation after the preparation of the stimulated resin root canal with two single-file systems, namely Waveone and Reciproc. Materials and Methods: Thirty stimulated resin root canal blocks with size 8/0. 02 K file were randomly divided into two study groups. The preparation in Group A and Group B was performed using Reciproc and Waveone files, respectively. Pre and post- preparation photographs were taken and the images were superimposed to evaluate the inner and outer wall’s curvature tendency at three points (apical, middle and coronal using AutoCad pragram. Data were analyzed using T-test. Results: Based on the results, the degree of transportation in the inner and outer walls of the canal was less at the level of 3 millimeters (P0.05. Conclusion: Waveone showed better performance in the middle third of canal and this system maybe recommended.

  7. A TRANSPORTATION RISK ASSESSMENT TOOL FOR ANALYZING THE TRANSPORT OF SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE TO THE PROPOSED YUCCA MOUNTAIN REPOSITORY

    International Nuclear Information System (INIS)

    2001-01-01

    The Yucca Mountain Draft Environmental Impact Statement (DEIS) analysis addressed the potential for transporting spent nuclear fuel and high-level radioactive waste from 77 origins for 34 types of spent fuel and high-level radioactive waste, 49,914 legal weight truck shipments, and 10,911 rail shipments. The analysis evaluated transportation over 59,250 unique shipment links for travel outside Nevada (shipment segments in urban, suburban or rural zones by state), and 22,611 links in Nevada. In addition, the analysis modeled the behavior of 41 isotopes, 1091 source terms, and used 8850 food transfer factors (distinct factors by isotope for each state). The analysis also used mode-specific accident rates for legal weight truck, rail, and heavy haul truck by state, and barge by waterway. This complex mix of data and information required an innovative approach to assess the transportation impacts. The approach employed a Microsoft(reg s ign) Access database tool that incorporated data from many sources, including unit risk factors calculated using the RADTRAN IV transportation risk assessment computer program. Using Microsoft(reg s ign) Access, the analysts organized data (such as state-specific accident and fatality rates) into tables and developed queries to obtain the overall transportation impacts. Queries are instructions to the database describing how to use data contained in the database tables. While a query might be applied to thousands of table entries, there is only one sequence of queries that is used to calculate a particular transportation impact. For example, the incident-free dose to off-link populations in a state is calculated by a query that uses route segment lengths for each route in a state that could be used by shipments, populations for each segment, number of shipments on each segment, and an incident-free unit risk factor calculated using RADTRAN IV. In addition to providing a method for using large volumes of data in the calculations, the

  8. Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix C, marine transport and associated environmental impacts. Volume 2

    International Nuclear Information System (INIS)

    1995-03-01

    This is Appendix C to a Draft Environmental Statement on a Proposed Nuclear Weapon Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. Shipment of any material via ocean transport entails risks to both the ship's crew and the environment. The risks result directly from transportation-related accidents and, in the case of radioactive or other hazardous materials, also include exposure to the effects of the material itself. This appendix provides a description of the approach used to assess the risks associated with the transport of foreign research reactor spent nuclear fuel from a foreign port to a U.S. port(s) of entry. This appendix also includes a discussion of the shipping configuration of the foreign research reactor spent nuclear fuel, the possible types of vessels that could be used to make the shipments, the risk assessment methodology (addressing both incident-free and accident risks), and the results of the analyses. Analysis of activities in the port(s) is described in Appendix D. The incident-free and accident risk assessment results are presented in terms of the per shipment risk and total risks associated with the basic implementation of Management Alternative 1and other implementation alternatives. In addition, annual risks from incident-free transport are developed

  9. Conditioning of spent fuel assemblies from the Rossendorf RFR research reactor in transport and storage containers of the type CASTOR MTR 2

    International Nuclear Information System (INIS)

    Schneider, B.; Hofmann, G.

    1994-09-01

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

  10. BWR spent fuel transport and storage system for KKL: TN trademark 52L, TN trademark 97L, TN trademark 24 BHL

    International Nuclear Information System (INIS)

    Sicard, D.; Verdier, A.; Monsigny, P.A.

    2004-01-01

    The LEIBSTADT (KKL) nuclear power plant in Switzerland has opted to ship spent fuel to a central facility called ZWILAG for interim storage. In the mid-nineties, COGEMA LOGISTICS was contracted by KKL for the supply of the TN trademark a52L and TN trademark 97L transport and storage casks for BWR fuel types. In 2003, KKL also ordered from COGEMA LOGISTICS the supply of six TNae24 BHL transport and storage casks. This paper shows how all the three cask designs have responded to the KKL needs to ship and store BWR spent fuel. In addition, it highlights the already significant operational feedback of the TN trademark 52L and TN trademark 97L casks by the KKL and ZWILAG operators

  11. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  12. Determination of 63Ni and 59Ni in spent ion-exchange resin and activated charcoal from the IEA-R1 nuclear research reactor

    International Nuclear Information System (INIS)

    Taddei, M.H.T.; Macacini, J.F.; Vicente, R.; Marumo, J.T.; Sakata, S.K.; Terremoto, L.A.A.

    2013-01-01

    A radiochemical method has been adapted to determine 59 Ni and 63 Ni in samples of radioactive wastes from the water cleanup system of the IEA-R1 nuclear research reactor. The process includes extraction chromatographic resin with dimethylglyoxime (DMG) as a functional group. Activity concentrations of 59 Ni and 63 Ni were measured, respectively, by X-ray spectrometry and liquid scintillation counting, whereas the chemical yield was determined by ICP-OES. The average ratio of measured activity concentrations of 63 Ni and 59 Ni agree well with theory. - Highlights: ► A radiochemical method was adapted for determination of 59 Ni and 63 Ni. ► 59 Ni and 63 Ni were determined in samples of radioactive wastes from research reactor. ► The activity concentration of 59 Ni was measured by X-ray spectrometry. ► The activity concentration of 63 Ni was measured by liquid scintillation counting. ► Average ratio of measured activity concentrations of 63 Ni and 59 Ni agrees with theory

  13. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

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

    1997-12-01

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

  14. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

  15. Comparison of apical transportation and change of working length in K3, NRT AND PROFILE rotary instruments using transparent resin block

    Directory of Open Access Journals (Sweden)

    Min-Jung Yoon

    2011-01-01

    Full Text Available Objectives The purpose of this study is to compare the apical transportation and working length change in curved root canals created in resin blocks, using 3 geometrically different types of Ni-Ti files, K3, NRT, and Profile. Materials and Methods The curvature of 30 resin blocks was measured by Schneider technique and each groups of Ni-Ti files were allocated with 10 resin blocks at random. The canals were shaped with Ni-Ti files by Crown-down technique. It was analyzed by Double radiograph superimposition method (Backman CA 1992, and for the accuracy and consistency, specially designed jig, digital X-ray, and CAD/CAM software for measurement of apical transportation were used. The amount of apical transportation was measured at 0, 1, 3, 5 mm from 'apical foramen - 0.5 mm' area, and the alteration of the working length before and after canal shaping was also measured. For statistics, Kruskal-Wallis One Way Analysis was used. Results There was no significant difference between the groups in the amount of working length change and apical transportation at 0, 1, and 3 mm area (p = 0.027, however, the amount of apical transportation at 5 mm area showed significant difference between K3 and Profile system (p = 0.924. Conclusions As a result of this study, the 3 geometrically different Ni-Ti files showed no significant difference in apical transportation and working length change and maintained the original root canal shape.

  16. Regular Advisory Group on Spent Fuel Management

    International Nuclear Information System (INIS)

    1993-01-01

    The Regular Advisory Group on Spent Fuel Management (RAGSFM) was established in accordance with the recommendations of the Expert Group on International Spent Fuel Management in 1982. The Advisory Group consists of nominated experts from countries with considerable experience and/or requirements in such aspects of the back-end of the fuel cycle as storage, safety, transportation and treatment of spent fuel. The RAGSFM activities cover the following main topics: a) Analysis and summary of spent fuel arisings and storage facilities; b) Interface between spent fuel storage and transportation activities; c) Spent fuel storage process and technology and related safety issues; d)Treatment of spent fuel

  17. The utility industry's perspective on OCRWM's plans for developing the system for transporting spent fuel under the Nuclear Waste Policy Act

    International Nuclear Information System (INIS)

    Brodnick, D.A.

    1988-01-01

    The electric utility industry has a vital interest in the transport program to be developed by the Department of Energy's Office of Civilian Radioactive Waste Management under the Nuclear Waste Policy Act. The industry's interest stems in part from the fact that the DOE's transportation program is financed by the Nuclear Waste Fund which is made up of ratepayer funds. However, the industry is also vitally interested in the DOE's transportation program because it could impact the ongoing transportation operations of all nuclear utilities, and, perhaps most importantly, without the utility industry's input, DOE is not able to develop an optimal transportation program. The NWPA contemplates that the DOE conducts its transportation program in accordance with the existing federal and state regulatory structure. DOE has significant discretion, however, in creating and implementing the business, operational and institutional aspects of its NWPA transportation program. The utility industry intends to ensure that the DOE meets the challenge to develop a safe, efficient and economically sound program to transport spent fuel and high-level waste to the appropriate federal facilities

  18. The Development of an Effective Transportation Risk Assessment Model for Analyzing the Transport of Spent Fuel and High-Level Radioactive Waste to the Proposed Yucca Mountain Repository

    International Nuclear Information System (INIS)

    McSweeney, Thomas; Winnard, Thomas; Ross, Steven B.; Best, Ralph E.

    2001-01-01

    Past approaches for assessing the impacts of transporting spent fuel and high-level radioactive waste have not been effectively implemented or have used relatively simple approaches. The Yucca Mountain Draft Environmental Impact Statement (DEIS) analysis considers 83 origins, 34 fuel types, 49,914 legal weight truck shipments, 10,911 rail shipments, consisting of 59,250 shipment links outside Nevada (shipment kilometers and population density pairs through urban, suburban or rural zones by state), and 22,611 shipment links in Nevada. There was additional complexity within the analysis. The analysis modeled the behavior of 41 isotopes, 1091 source terms, and used 8850 food transfer factors (distinct factors by isotope for each state). The model also considered different accident rates for legal weight truck, rail, and heavy haul truck by state, and barge by waterway. To capture the all of the complexities of the transportation analysis, a Microsoft(reg s ign) Access database was created. In the Microsoft(reg s ign) Access approach the data is placed in individual tables and equations are developed in queries to obtain the overall impacts. While the query might be applied to thousands of table entries, there is only one equation for a particular impact. This greatly simplifies the validation effort. Furthermore, in Access, data in tables can be linked automatically using query joins. Another advantage built into MS Access is nested queries, or the ability to develop query hierarchies. It is possible to separate the calculation into a series of steps, each step represented by a query. For example, the first query might calculate the number of shipment kilometers traveled through urban, rural and suburban zones for all states. Subsequent queries could join the shipment kilometers query results with another table containing the state and mode specific accident rate to produce accidents by state. One of the biggest advantages of the nested queries is in validation

  19. Method of separating radioactive nuclides from ion exchange resins

    International Nuclear Information System (INIS)

    Suzuki, Kazunori; Saikoku, Masami; Taneta, Daisuke; Yagi, Takuro.

    1987-01-01

    Purpose: To enable to safely process radioactive nuclides from spent ion exchange resins by using existent processing facilities. Method: Ion exchange resins in aqueous medium are at first placed to the ultrasonic wave irradiation site and put into such a state where clads and resins are easily separatable from each other by weakening the bonding force between them. Since the clads are magnetic material such as Fe 3 O 4 or NiFe 2 O 4 , the clads can be collected in the direction of the magnetic force by exerting the magnetic field simultaneously. The collected clads are transported by means of the aqueous medium to a collecting tank by removing the effect of magnetic field, for example, by interrupting the current supply to the electromagnet. Finally, they were subjected to stabilization and fixation into inorganic hardening agent such as cement hardener. Thus, processions can be made safely by using existent facilities. (Takahashi, M.)

  20. Test program of the drop tests with full scale and 1/2.5 scale models of spent nuclear fuel transport and storage cask

    International Nuclear Information System (INIS)

    Kuri, S.; Matsuoka, T.; Kishimoto, J.; Ishiko, D.; Saito, Y.; Kimura, T.

    2004-01-01

    MHI have been developing 5 types of spent nuclear fuel transport and storage cask (MSF cask fleet) as a cask line-up. In order to demonstrate their safety, a representative cask model for the cask fleet have been designed for drop test regulated in IAEA TS-R-1. The drop test with a full and a 1/2.5 scale models are to be performed. It describes the test program of the drop test and manufacturing process of the scale models used for the tests

  1. Analysis of radiation doses from operation of postulated commercial spent fuel transportation systems: Analysis of a system containing a monitored retrievable storage facility

    International Nuclear Information System (INIS)

    Smith, R.I.; Daling, P.M.; Faletti, D.W.

    1992-04-01

    This addendum report extends the original study of the estimated radiation doses to the public and to workers resulting from transporting spent nuclear fuel from commercial nuclear power reactor stations through the federal waste management system (FWMS), to a system that contains a monitored retrievable storage (MRS) facility. The system concepts and designs utilized herein are consistent with those used in the original study (circa 1985--1987). Because the FWMS design is still evolving, the results of these analyses may no longer apply to the design for casks and cask handling systems that are currently being considered. Four system scenarios are examined and compared with the reference No-MRS scenario (all spent fuel transported directly from the reactors to the western repository in standard-capacity truck and rail casks). In Scenarios 1 and 2, an MRS facility is located in eastern United States and ships either intact fuel assemblies or consolidated fuel rods and compacted assembly hardware in canisters. In Scenarios 3 and 4, an MRS facility is located in the western United States and ship either intact fuel assemblies or consolidated fuel rods and compacted assembly hardware in canisters

  2. The so-called 'Castor-Crisis': Transport of spent nuclear fuel elements and German 'Angst'. How to prevent the public relations catastrophe

    International Nuclear Information System (INIS)

    Suess, Werner

    1999-01-01

    Full text: 1. 'Castor-Crisis' - The Real Facts. - the background: radioactive contamination on the surface of transport containers for spent nuclear fuel elements; - legal aspects: transport limit values and notification obligations; - health aspects: radioactive contamination and ionising radiation; - the news media: divergence between technical facts and public perception. 2. 'Castor-Crisis' - The Reactions. 2.1 Technical measures: - 'action plan' of the Federal Ministry for Environment Protection and Reactor Safety; - IT-based European Information and Report System for the transport of nuclear combustibles => 'Transparent Transport Procedures'; - optimisation of decontamination procedures and transport organisation; - simplification of logistics, clearer responsibilities. 2.2 Communications measures: - defense strategy: 'we made a mistake...'; - information campaign: 'we owe you some answers...'; - regaining credibility: public testimonials of employees in newspaper ads, brochures etc.; - regaining credibility: neutral investigation of all relevant circumstances by KPMG. 3. 'Castor-Crisis' - The Lessons: - internal crisis management: improved co-ordination at company and branch level; - pro-active strategy: 'The benefits of nuclear energy' (avoidance Of CO 2 -emissions); - limits of communications; - communications efforts for nuclear energy - the European context. (author)

  3. Investigation of nuclide importance to functional requirements related to transport and long-term storage of LWR spent fuel

    International Nuclear Information System (INIS)

    Broadhead, B.L.; DeHart, M.D.; Ryman, J.C.; Tang, J.S.; Parks, C.V.

    1995-06-01

    This study investigates the relative importances of the various actinide, fission-product, and light-element isotopes associated with LWR spent fuel with respect to five analysis areas: criticality safety (absorption fractions), shielding (dose rate fractions), curies (fractional curies levels), decay heat (fraction of total watts), and radiological toxicity (fraction of potential committed effective dose equivalent). These rankings are presented for up to six different burnup/enrichment scenarios and at decay times from 2 to 100,000 years. Ranking plots for each of these analysis areas are given in an Appendix for completeness, as well as summary tables in the main body of the report. Summary rankings are presented in terms of high (greater than 10% contribution to the total), medium (between 1% and 10% contribution), and low (less than 1% contribution) for both short- and long-term cooling. When compared with the expected measurement accuracies, these rankings show that most of the important isotopes can be characterized sufficiently for the purpose of radionuclide generation/depletion code validation in each of the analysis areas. Because the main focus of this work is on the relative importances of isotopes associated with L at sign spent fuel, some conclusions may not be applicable to similar areas such as high-level waste (HLW) and nonfuel-bearing components (NFBC)

  4. Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

    2007-10-01

    The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department of Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial reactor

  5. Calculation of radiation doses and thermal loads for specific situations that may occur during the transport and storage of spent fuel assemblies from the Rossendorf reactor

    International Nuclear Information System (INIS)

    Seifert, E.

    1995-01-01

    The CAeSAR-III transfer cask and the CASTOR MTR 2 transport and storage cask have been selected for accommodating the spent fuel assemblies from the Rossendorf reactor. The fuel assemblies are packed into the loading units consisting of Al cylinders provided with the suitable gaps for insertion of the spent fuel, according to the various types of fuel elements (EK10, WWR-M, WWR-M2, tubular elements). From among the various types of loading units, only type A designed for accepting 21 single fuel assemblies ( or 7 triple elements) of WWR-M or WWR-M2 is of significance in this context, as they represent the covering neutron and gamma source taken as the radiation source for the calculations. The CAeSAR cask can accommodate one of these loading units, and the CASTOR seven. The dose calculations performed with the MCNP Monte-Carlo code (equivalent dose rate) refer to the loaded casks and some hypothetical situations occurring during loading. The thermal load calculations include the heat-up curve in the event of ''Fire'' in the CASTOR cask, and the stationary temperature distribution in the CASTOR cask due to internal heat generation ( darioactive decay of the inserted fuel assemblies). The HEATING program is used for this purpose. (orig.) [de

  6. Shielding analysis of a transport and storage cask for spent BWR fuel applicability of the code SAS4 and discussion of results

    Energy Technology Data Exchange (ETDEWEB)

    Hilbert, F. [Nuclear Cargo and Service GmbH, Hanau (Germany); Morishima, M.; Tamaki, H. [Mitsubishi Heavy Industries, Kobe (Japan)

    2004-07-01

    For the shielding analysis of transport and storage casks for spent fuel the use of computer codes is state of the art. However, in most applications the computer models used for the analysis are simplified to circular geometries to save modelling effort and calculation time. Furthermore, the active zone of the fuel is modelled as homogeneous zone with a uniform average burn-up. In the first part of the present paper it is shown that an exact model is feasible and the effect of the geometrical shape on the dose rates is illustrated. The second part of the paper shows the comparison of the dose rates calculated with 5 different fuel models. Finally, the accuracy of the calculations is discussed.

  7. Reference concept for the direct final disposal of spent HTR-FEs in CASTOR THTR/AVR transport and storage casks

    International Nuclear Information System (INIS)

    Niephaus, D.

    2000-01-01

    For the final disposal of spent HTR-FEs from the two decommissioned THTR 300 and AVR high-temperature reactor plants, which have been packaged in CASTOR THTR/AVR transport and storage casks for long-term storage in the Ahaus interim storage facility and the AVR store at Juelich, an appropriate horizontal drift emplacement concept based on the CASTOR THTR/AVR transport and storage cask is developed and presented in this report. First of all, the essential design and conceptual features of the AVR and THTR fuel elements will be described and the HTR-FE quantity structure for final disposal compiled. Furthermore, the findings and experience gained experimentally and by safety-related studies and calculations as well as, in particular, the results and experience obtained for the borehole storage concept by long-term safety analyses concerning the storage behaviour of spent HTR-FEs will be described in detail and used as a yardstick for the assessment of the long-term safety of the horizontal drift emplacement concept on the basis of the CASTOR THTR/AVR cask. For the accident of brine in leakage into the abandoned underground workings assumed in the long-term safety analyses it will be shown by stress and deformation calculations that the CASTOR THTR/AVR casks on the whole have sufficient safety reserves to sustain the tectonic loads to be expected and that a lifetime is to be expected which is currently also taken as a standard for POLLUX casks. (orig.) [de

  8. Spent-fuel-storage alternatives

    International Nuclear Information System (INIS)

    1980-01-01

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

  9. Spent fuel management in Japan

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  10. ALARA Principle Application for Loading Spent Nuclear Fuel Assemblies from Nuclear Research Reactor WR-S Mergal-Bucharest Romania into Transportation Casks

    International Nuclear Information System (INIS)

    Dragusin, M.

    2009-01-01

    Safety implementation of Spent Nuclear Fuels Assemblies (SNFA) handling procedures at the WR-S reactor site is ensured by technical perfection and reliability of equipment, monitoring of its condition, qualification and discipline of personnel as well as organization and execution of work complied with requirements of regulatory documents, process procedures, guidance and manuals. The personnel training for execution loading of SNF FAs is other important aspect for radiation protection and safely activities. Estimations carried out using Micro Shield software show that maximal dose rate upon working site when loading four FAs into basket of cask will not exceed 1.7 and 956;Sv/h, excluding natural radiation. Radiation Safety Analyses estimates for loading 70 SNFA in 18 transportation casks are: maximal individual dose: 4274.7 and 956;Sv, maximal expected collective dose persons: 17 031.2 man and 956;Sv. By application ALARA principle with technical and administrative measures the loading process developed in the following conditions: maximal individual dose: 68 and 956;Sv, the collective dose persons: 732 man and 956;Sv. The work will presented the technical measures and procedures applied in loading process.

  11. Near surface spent fuel storage: environmental issues

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  12. Draft environmental impact statement on a proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Volume 2, Appendix E, Evaluation of human health effects of overland transportation

    International Nuclear Information System (INIS)

    1995-03-01

    This Appendix provides an overview of the approach used to assess the human health risks that may result from the overland transportation of foreign research reactor spent nuclear fuel. The Appendix includes discussion of the scope of the assessment, analytical methods used for the risk assessment (i.e., computer models), important assessment assumptions, determination of potential transportation routes, and presents the results of the assessment. In addition, to aid in the understanding and interpretation of the results, specific arm of uncertainty are described, with an emphasis an how the uncertainties may affect comparisons of the alternatives. he approach used in this Appendix is modeled after that used in the Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Draft Environmental Impact Statement (SNF ampersand INEL Draft EIS) (DOE, 1994b). The SNF ampersand INEL Draft EIS did not perform as detailed an analysis on the specific actions taken for foreign research reactor spent nuclear fuel because of the breadth necessary to analyze the entire spent fuel management program. However, the fundamental assumptions used in this analysis are consistent with those used in the SNF ampersand INEL Draft EIS (DOE, 1994b), and the same computer codes and generic release and accident data are used. The risk assessment results are presented in this Appendix in terms of ''Per-shipment'' risk factors, as well as for the total risks associated with each alternative. Per-shipment risk factors provide an estimate of the risk from a single spent nuclear fuel shipment between a specific origin and destination. They are calculated for all possible origin and destination pairs for each spent nuclear fuel type. The total risks for a given alternative are found by multiplying the expected number of shipments by the appropriate per-shipment risk factors. This approach provides maximum flexibility for determining the risks for a large number of potential

  13. Safety analysis of spent fuel packaging

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

    1987-01-01

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

  14. Potential problems associated with ion-exchange resins used in the decontamination of light-water reactor systems

    International Nuclear Information System (INIS)

    Soo, P.; Adams, J.W.; Kempf, C.R.

    1987-01-01

    During a typical decontamination event, ion-exchange resin beds are used to remove corrosion products (radioactive and nonradioactive) and excess decontamination reagents from waste streams. The spent resins may be solidified in a binder, such as cement, or sealed in a high-integrity container (HIC) in order to meet waste stability requirements specified by the Nuclear Regulatory Commission. Lack of stability of low-level waste in a shallow land burial trench may lead to trench subsidence, enhanced water infiltration and waste leaching, which would result in accelerated transport of radionuclides and the complexing agents used for decontamination. The current program is directed at investigating safety problems associated with the handling, solidification and containerization of decontamination resin wastes. The three tasks currently underway include freeze-thaw cycling of cementitious and vinyl ester-styrene forms to determine if mechanical integrity is compromised, a study of the corrosion of container materials by spent decontamination waste resins, and investigations of resin degradation mechanisms

  15. Modelling of radionuclide transport in crystalline basement for demonstration of spent nuclear fuel disposal feasibility (case of South-eastern Lithuania)

    International Nuclear Information System (INIS)

    Jakimaviciute-Maseliene, V.; Mazeika, J.; Petrosius, R.

    2006-01-01

    The Strategy of Radioactive Waste Management of Lithuania provides for evaluating the possibilities of disposal in a deep geological repository of spent nuclear fuel and long-lived radioactive waste originated from the Ignalina NPP. The initial feasibility studies performed in Lithuania during 2001-2004 focus on screening all potentially prospective rock types as a host medium for a repository. Since most information is available on crystalline basement and sedimentary cover of South-eastern Lithuania, this host medium was selected for model case studies. Taking into account the main assumptions (the canister defect scenario proposed by Swedish experts and evaluated by Lithuanian Energy Institute experts) groundwater flow and radionuclide (iodine-129 as a mobile and long-lived one) transport modeling using FEFLOW computer code was performed to model the domain (approximately a 0.8 km long, 0.6 km wide and 0.52 km thick far-field block) of South-eastern Lithuania. The model domain comprises a Proterozoic-Archaean aquifer with an overlaying aquifer system of sedimentary cover. The upward groundwater flow through a defected canister located in a tectonically damaged zone was conservatively generated. The main results of calculations are the following: in case of upward groundwater flow, the maximum activity concentration of I-129 in the single longitudinal tectonic fracture above the defected canister will not exceed 10 -4 Bq/l, and in the active water exchange zone it is close to 10 -7 Bq/l. These figures show that doses obtained by a human recipient via the aquatic pathway should be below the dose limit (1 mSv/y). Location of the model domain in Southeastern Lithuania does not mean any reference to the site for a deep geological repository. (author)

  16. Status and prospects for spent fuel management in France

    International Nuclear Information System (INIS)

    Portal, R.; L'Epine, P. de

    1996-01-01

    The spent fuel arisings and storage capacities, the interface between fuel storage and transportation activities, the spent fuel storage technology, the reprocessing and recycling industrial activities in France are described in the paper. (author). 6 figs, 8 tabs

  17. Spent-fuel-storage alternatives

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

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

  18. Effects of environments on spent fuel

    International Nuclear Information System (INIS)

    Funk, C.W.; Jacobson, L.D.; Menon, M.N.

    1979-07-01

    This report describes the influence of water storage environment and transportation on spent light water reactor (LWR) fuel assemblies. It also estimates the storage duration and capacity requirements for several assumed scenarios

  19. Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

    Energy Technology Data Exchange (ETDEWEB)

    Cuta, Judith M.; Jenquin, Urban P.; McKinnon, Mikal A.

    2001-11-20

    The licensing of spent fuel storage casks is generally based on conservative analyses that assume a storage system being uniformly loaded with design basis fuel. The design basis fuel typically assumes a maximum assembly enrichment, maximum burn up, and minimum cooling time. These conditions set the maximum decay heat loads and radioactive source terms for the design. Recognizing that reactor spent fuel pools hold spent fuel with an array of initial enrichments, burners, and cooling times, this study was performed to evaluate the effect of load pattern on peak cladding temperature and cask surface dose rate. Based on the analysis, the authors concluded that load patterns could be used to reduce peak cladding temperatures in a cask without adversely impacting the surface dose rates.

  20. Immobilisation of ion exchange resins in cement

    International Nuclear Information System (INIS)

    Howard, C.G.; Jolliffe, C.B.; Lee, D.J.

    1990-09-01

    The removal of activity from spent decontaminating solutions eg LOMI can be achieved using organic ion exchange resins. These resins can be successfully immobilised in cement based matrices. The optimum cement system contained 10% ordinary Portland cement 84% gg blast furnace slag, 6% microsilica with a water cement ratio of 0.5 and a dry resin loading of 36% with respect to total weight. This formulation was successfully scaled up to 200 litres giving a product with acceptable compressive strength, dimensional stability and elastic modulus. Storage of samples under water appears to have no detrimental effects on the product's properties. (author)

  1. Plastic casting resin poisoning

    Science.gov (United States)

    Epoxy poisoning; Resin poisoning ... Epoxy and resin can be poisonous if they are swallowed or their fumes are breathed in. ... Plastic casting resins are found in various plastic casting resin products.

  2. Permanganate Degradation of Reillex HPQ Ion Exchange Resin for Use in HB-Line

    International Nuclear Information System (INIS)

    Walker, B.W.

    1999-01-01

    This study evaluated the use of Reillex TM HPQ resin as a replacement for the Ionac A-641 resin currently authorized for use in H B-Line. The study concentrated on the ability of the existing alkaline permanganate digestion process to convert spent resin for disposal

  3. Intermodal transfer of spent fuel

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  4. Leaching of 137Cs from the ion-exchange resin incorporated in polyethylene or cement composite

    International Nuclear Information System (INIS)

    Mariyama, N.; Dojiri, S.; Matsuzuru, H.

    1977-01-01

    The results of an evaporation of a composite, which has been developed by incorporation of spent ion-exchange resin in polyethylene for the solidification of radioactive wastes, are reported. Transport phenomena involved in the leaching of 137 Cs from the composite matrix into surrounding water were investigated using two methods based on theoretical equations, a diffusion equation derived for a plane source model, and a rate equation for diffusion coupled with a first-order reaction. The leaching data were also analyzed by an empirical method employing a polynomial equation. Comparative results are presented. (U.K.)

  5. Resin composites

    DEFF Research Database (Denmark)

    Benetti, Ana Raquel; Peutzfeldt, Anne; Lussi, Adrian

    2014-01-01

    OBJECTIVE: To investigate how the modulus of elasticity of resin composites influences marginal quality in restorations submitted to thermocyclic and mechanical loading. METHODS: Charisma, Filtek Supreme XTE and Grandio were selected as they were found to possess different moduli of elasticity...... of resin composite (p=0.81) on the quality of dentine margins was observed, before or after loading. Deterioration of all margins was evident after loading (p....008). CONCLUSIONS: The resin composite with the highest modulus of elasticity resulted in the highest number of gap-free enamel margins but with an increased incidence of paramarginal enamel fractures. CLINICAL SIGNIFICANCE: The results from this study suggest that the marginal quality of restorations can...

  6. 5 CFR 551.422 - Time spent traveling.

    Science.gov (United States)

    2010-01-01

    ... 5 Administrative Personnel 1 2010-01-01 2010-01-01 false Time spent traveling. 551.422 Section 551... Activities § 551.422 Time spent traveling. (a) Time spent traveling shall be considered hours of work if: (1... who is permitted to use an alternative mode of transportation, or an employee who travels at a time...

  7. Validation of computer codes and modelling methods for giving proof of nuclear saefty of transport and storage of spent VVER-type nuclear fuels. Part 1. Purposes and goals of the project. Final report

    International Nuclear Information System (INIS)

    Buechse, H.; Langowski, A.; Lein, M.; Nagel, R.; Schmidt, H.; Stammel, M.

    1995-01-01

    The report gives the results of investigations on the validation of computer codes used to prove nuclear safety during transport and storage of spent VVER - fuel of NPP Greifswald and Rheinsberg. Characteristics of typical spent fuel (nuclide concentration, neutron source strength, gamma spectrum, decay heat) - calculated with several codes - and dose rates (e.g. in the surrounding of a loaded spent fuel cask) - based on the different source terms - are presented. Differences and their possible reasons are discussed. The results show that despite the differences in the source terms all relevant health physics requirements are met for all cases of source term. The validation of the criticality code OMEGA was established by calculation of appr. 200 critical experiments of LWR fuel, including VVER fuel rod arrangements. The mean error of the effective multiplication factor k eff is -0,01 compared to the experiment for this area of applicability. Thus, the OMEGA error of 2% assumed in earlier works has turned out to be sufficiently conservative. (orig.) [de

  8. Time/motion observations and dose analysis of reactor loading, transportation, and dry unloading of an overweight truck spent fuel shipment

    International Nuclear Information System (INIS)

    Hostick, C.J.; Lavender, J.C.; Wakeman, B.H.

    1992-04-01

    This document presents observed activity durations and radiation dose analyses for an overweight truck shipment of pressurized water reactor (PWR) spent fuel from the Surry Power Station in Virginia to the Idaho National Engineering Laboratory. The shipment consisted of a TN-8L shipping cask carrying three 9-year-old PWR spent fuel assemblies. Handling times and dose analyses for at-reactor activities were completed by Virginia Electric and Power Company (Virginia Power) personnel. Observations of in-transit and unloading activities were made by Pacific Northwest Laboratory (PNL) personnel, who followed the shipment for approximately 2800 miles and observed cask unloading activities. In-transit dose estimates were calculated using dose rate maps provided by Virginia Power for a fully loaded TN-8L shipping cask. The dose analysis for the cask unloading operations is based on the observations of PNL personnel

  9. Past and future IAEA spent fuel management activities

    International Nuclear Information System (INIS)

    Grigoriev, A.

    1993-01-01

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

  10. Spent nuclear fuel project product specification

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    1999-01-01

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project

  11. Current status of spent fuel management in China

    International Nuclear Information System (INIS)

    Liu Yuansong

    2005-01-01

    The administration organization and the laws and regulations on spent fuel management in China are described at first. In the year of 2000, CNN C and GNPJVC signed a contract of take-over mode for spent fuel, in which CNNC will take-over the spent fuel annually since 2003. In the year of 2001, CNNC and QNPC signed another contract of take-over mode for spent fuel. Transportation by road is the realistic selections of the transportation of the spent fuel from the nuclear power stations to LNFC. As for spent fuel management, the construction of CNNC's reprocessing pilot plant is going on and R and D of spent fuel management are launching continuously. (author)

  12. A preliminary assessment of system cost impacts of using transportable storage casks and other shippable metal casks in the utility/DOE spent fuel management system

    International Nuclear Information System (INIS)

    Johnson, E.R.

    1988-01-01

    In view of the foregoing, a study was conducted by E.R. Johnson Associates, Inc. and H and R Technical Associates, Inc. to determine the prospective viability of the use of TSCs and shippable SOCs in the combined utility/DOE system. This study considered costs, ALARA considerations and the logistics of the use and delivery of casks to the DOE system by utilities. It was intended that this study would result in a technical and cost resource base that could be used for evaluating various strategies and scenarios for deploying TSCs or SOCs in the combined utility/DOE spent fuel management system with respect to the prospective economic advantage that could be realized

  13. Review: Resin Composite Filling

    OpenAIRE

    Desmond Ng; Jimmy C. M. Hsiao; Keith C. T. Tong; Harry Kim; Yanjie Mai; Keith H. S. Chan

    2010-01-01

    The leading cause of oral pain and tooth loss is from caries and their treatment include restoration using amalgam, resin, porcelain and gold, endodontic therapy and extraction. Resin composite restorations have grown popular over the last half a century because it can take shades more similar to enamel. Here, we discuss the history and use of resin, comparison between amalgam and resin, clinical procedures involved and finishing and polishing techniques for resin restoration. Although resin ...

  14. The development of an operations system for the transport of spent nuclear fuel in the United States Civilian Radioactive Waste Management Program

    International Nuclear Information System (INIS)

    Best, R.E.; Danese, F.L.; Peterson, R.W.; Joy, D.S.; Pope, R.B.; Ratledge, J.E.; Shappert, L.B.; Wankerl, M.W.; Klimas, M.J.; Darrough, M.E.

    1990-01-01

    In order to support the development of a Transportation Operations System for the Federal Waste Management System (FWMS) by the Office of Civilian Radioactive Waste Management, the United States Department of Energy formed the Transportation Project Office at its field office in Chicago. Planning and development activities are being performed in a number of areas including a major effort in operations support, providing the planning and assessment necessary for developing the future transportation operations capability needed by the FWMS. The purpose of this paper is to review significant planning and development accomplishments, and outline expected future efforts for the continued development, acquisition, test, and startup of the transportation operations component of the FWMS. 2 refs

  15. Carbon-14 removal for disposal of reactor deionizer resins

    International Nuclear Information System (INIS)

    Carlton, W.H.; Baumann, E.W.

    1991-01-01

    Disposal of depleted ion exchange resins from the primary system of the Savannah River Site (SRS) reactors is complicated by the presence of Carbon-14. Because Carbon-14 has a long half-life (5,730 years) and high mobility in soils, burial of the resins is no longer a viable option. Consequently some 35 spent reactor deionizers have accumulated that are to be stored aboveground in H-Area for an indefinite period. Spent deionizers containing Carbon-14 will continue to accumulate with operation of the present production reactors and would also accumulate from the proposed heavy water new production reactor. Removal of the Carbon-14 from the resins would reduce the volume of Carbon-14 bearing waste and enable the resins to be disposed of as low-level waste. Studies at SRS have indicated that the Carbon-14 from reactor primary coolant is mostly retained by the resins as the bicarbonate anion. Thus Carbon-14 removal might be accomplished by an acidification operation with trapping of the carbon dioxide released, for separate disposal. Conversion of the bicarbonate from the resin to barium carbonate, for example, would reduce the volume of waste more than a hundredfold. Displacement and recovery of Carbon-14 dioxide from reactor coolant deionizers by acid treatment has been reported by the Canadians. This memorandum recommends that a process be developed for Carbon-14 dioxide removal from SRS spent reactor deionizer resins, drawing on the Canadian experience

  16. Status of spent fuel shipping cask development

    International Nuclear Information System (INIS)

    Hall, I.K.; Hinschberger, S.T.

    1989-01-01

    This paper discusses how several new-generation shopping cask systems are being developed for safe and economical transport of commercial spent nuclear fuel and other radioactive wastes for the generating sites to a federal geologic repository or monitored retrievable storage (MRS) facility. Primary objectives of the from-reactor spent fuel cask development work are: to increase cask payloads by taking advantage of the increased at-reactor storage time under the current spent fuel management scenario, to facilitate more efficient cask handling operations with reduced occupational radiation exposure, and to promote standardization of the physical interfaces between casks and the shipping and receiving facilities. Increased cask payloads will significantly reduce the numbers of shipments, with corresponding reductions in transportation costs and risks to transportation workers, cask handling personnel, and the general public

  17. Resin Characterization

    Science.gov (United States)

    2015-06-01

    to see plastic deformation of the surface. 8.1.4.3 Density: Density using the Archimedes principle (ASTM D 792). 8.1.4.4 Density as a Function of...the cure and postcure, quickly cool the sample to 0 °C or lower the temperature to quench the reaction, and then ramp the temperature at 5 °C/min to...prepared by pouring 10 g of resin into a 30-mL screw-cap scintillation vial and adding appropriate amounts of initiator, catalyst, and inhibitor

  18. Processing of exhausted resins for Trino NPP,

    International Nuclear Information System (INIS)

    Benvenuto, F.; Bitetti, G.; Maggini, F.; Scarsi, G.

    2009-01-01

    Decomposition of organic compounds contained in the spent ion exchange resins is considered effective in reducing the waste volume. A system using the wet-oxidation process has been studied for the treatment of the spent resins stored at Trino Nuclear Power Plant owned by SOGIN. Compared with various processes for treating sludge and resin, the wet-oxidation system is rather simple and the process conditions are mild. Not contaminated ion exchange resin samples similar to those ones used in Trino NPP were processed by wet-oxidation and appropriate decomposition of the organic compounds was verified. After decomposition the residue can be solidified with cement for final disposal. When compared with direct solidification without decomposition, the number of waste packages can be significantly reduced. Additional measures for conditioning secondary waste products have also been studied, and their applicability to the Trino Nuclear Power Plant was verified. Some of conditions studied were specific to the Trino Nuclear Power Plant, but it is expected that the system will provide an effective solution for resin treatment at other Italian NPPs. (authors)

  19. Anion-exchange resin-based desulfurization process

    Energy Technology Data Exchange (ETDEWEB)

    Sheth, A.C.; Strevel, S.D.

    1991-01-01

    The University of Tennessee Space Institute (UTSI) has a Department of Energy grant to further develop the Institute's anion-exchange resin-based flue gas, desulfurization concept. The developmental program proposed includes screening of commercially available resins to select three candidate resins for further study. These three resins will undergo a series of experiments designed to test the resins' performance under different process conditions (including the use of spent MHD seed material). The best of these resins will be used in optimizing the regeneration step and in testing the effects of performance enhancers. The process schematic developed from the results will be used to estimate the related economics.

  20. Direct ultimate disposal of spent fuel. Simulation of shaft transport. Probabilistic safety analysis of a shaft hoisting equipment for a max. payload of 85 t (TA 11)

    International Nuclear Information System (INIS)

    Filbert, W.; Leicht, R.; Schaub, B.

    1994-03-01

    The reported PSA examined transport processes involved in the direct disposal of POLLUX containers in a radwaste repository. The processes analysed are loading of the hoisting cage above ground, shaft transport to the underground storage place, and discharge from the hoisting cage and emplacement of the container. The PSA results yield data defining the rate of occurrence of events described in the following, for an overall operating time of 10.000 transport processes, average duration of 30 minutes each. The events considered are: Class (1), (elevated radiation doses), probabilistic occurrence rate of 5.2 events per calendar year; Class (2), (release of radioactive materials), probabilistic occurrence rate of 1.33 x 10 -6 per calender year. These results are also applicable to the emplacement of other waste forms which are planned to be disposed of in the same radwaste site as the POLLUX containers. (orig./HP) [de

  1. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2005-01-01

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

  2. Forecasts and restrictions on vibrations from rock excavation and transportation. Encapsulation Plant and Repository for spent nuclear fuel, Laxemar; Prognoser och restriktioner foer vibrationer fraan bergschaktning och transporter. Inkapslingsanlaeggning och slutfoervar foer anvaent kaernbraensle, Laxemar

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Carl; Johansson, Sven-Erik (Nitro Consult AB (Sweden))

    2010-12-15

    This study describes the impact on the surroundings that may occur during rock excavation activities for the final repository for spent nuclear fuel in Laxemar and the encapsulation facility in Simpevarp. The study also includes vibrations created by heavy shipments related to activities at the final repository. The study will provide input to the environmental impact assessment and future design work. The survey area for buildings and facilities covered by the study extends approximately 1,000 metres from the proposed location of the final repository. For the encapsulation facility the survey area has been limited to residential buildings and summer houses within 1,000 metres of the proposed location. In addition, residential buildings along road 743 have been surveyed with regard to the impact of heavy shipments between Laxemar and Faarbo. The results of the surveys and information on planned rock excavation activities have been used to formulate preliminary restrictions and predictions of vibrations and air shock waves from blasting, as well as noise from rock drilling. Predictions have also been made of vibrations from heavy shipments, and a reference survey has been carried out in a residential building near road 743. The predictions of vibrations from blasting rounds reveal low or very low levels. No risk of damage to buildings or equipment is expected. Vibrations from blasting may, however, be perceptible within large parts of the study area, since the human perception threshold for vibration is very low. They will hardly be regarded as disturbing, however. When the accesses to the final repository have been built and rock excavation continues at repository level, the impact on the surroundings is expected to be minimal. The main reason for this is that the blasting will then occur at a depth of about 500 metres, at an ample distance to buildings at surface level. Predictions of air shock waves from blasting rounds indicate low levels. There is no risk of

  3. Spent nuclear fuel in Bulgaria

    International Nuclear Information System (INIS)

    Peev, P.; Kalimanov, N.

    1999-01-01

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

  4. The spent fuel fate

    International Nuclear Information System (INIS)

    2001-01-01

    The spent fuel is not a waste. It can be upgrade by a reprocessing which extracts all products able to produce energy. The today situation is presented and economically analyzed and future alternatives are discussed. (A.L.B.)

  5. Spent fuel workshop'2002

    International Nuclear Information System (INIS)

    Poinssot, Ch.

    2002-01-01

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

  6. Spent fuel centralized storage

    International Nuclear Information System (INIS)

    Chometon, P.L.

    1985-01-01

    Nuclear energy producer countries have felt the need to build a centralized spent fuel storage before reprocessing (for example, COGEMA in FRANCE), either in an adjoining plant on an appropriate site, or isolated. More rarely, this storage enables to decide whether to reprocess or to definitely store spent fuel considered as being waste: for example CLAB in Sweden. Our Company is specialized in the design and construction of spent fuel centralized storage plants. Storage generally takes place in a pool in order to facilitate handling operations and retrieving of these fuels, but these operations may also be effected in a dry way, either in concrete structures or in storage casks. With respect to pools, which might currently be the most appropriate and flexible system, several improvements have recently been made in the design of cask reception facilities and spent fuel storage. These improvements are presented, hereafter [fr

  7. Spent fuel centralized storage

    International Nuclear Information System (INIS)

    Baillif, L.; Chometon, P.L.

    1986-01-01

    Nuclear energy producer countries have felt the need to build a centralized spent fuel storage before reprocessing, either in an adjoining plant on an appropriate site, or isolated. More rarely, this storage enables to decide whether to reprocess or to definitely store spent fuel considered as being waste: for example CLAB in Sweden. Our Company SGN is specialized among others in the design and construction of spent fuel centralized storage plants. Storage generally takes place in a pool in order to facilitate handling operations and retrieving of these fuels, but these operations may also be effected in a dry way, either in concrete structures or in storage casks. With respect to pools, which might currently be the most appropriate and flexible system, several improvements have recently been made in the design of cask reception facilities and spent fuel storage. These improvements are presented, hereafter [fr

  8. Spent fuel management

    International Nuclear Information System (INIS)

    2005-01-01

    The production of nuclear electricity results in the generation of spent fuel that requires safe, secure and efficient management. Appropriate management of the resulting spent fuel is a key issue for the steady and sustainable growth of nuclear energy. Currently about 10,000 tonnes heavy metal (HM) of spent fuel are unloaded every year from nuclear power reactors worldwide, of which 8,500 t HM need to be stored (after accounting for reprocessed fuel). This is the largest continuous source of civil radioactive material generated, and needs to be managed appropriately. Member States have referred to storage periods of 100 years and even beyond, and as storage quantities and durations extend, new challenges arise in the institutional as well as in the technical area. The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs

  9. Development of advanced spent fuel management process. System analysis of advanced spent fuel management process

    International Nuclear Information System (INIS)

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

    1999-03-01

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

  10. Anion-exchange resin-based desulfurization process. Quarterly technical progress report, October 1, 1991--December 31, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Sheth, A.C.; Strevel, S.D.

    1991-12-31

    The University of Tennessee Space Institute (UTSI) has a Department of Energy grant to further develop the Institute`s anion-exchange resin-based flue gas, desulfurization concept. The developmental program proposed includes screening of commercially available resins to select three candidate resins for further study. These three resins will undergo a series of experiments designed to test the resins` performance under different process conditions (including the use of spent MHD seed material). The best of these resins will be used in optimizing the regeneration step and in testing the effects of performance enhancers. The process schematic developed from the results will be used to estimate the related economics.

  11. Spent Fuel Management Newsletter. No. 1

    International Nuclear Information System (INIS)

    1990-03-01

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

  12. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    Energy Technology Data Exchange (ETDEWEB)

    Il' kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation); Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K. [All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation); Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A. [All-Russian Research Inst. of Applied Chemistry, Moscow (Russian Federation); Haire, Jonathan M.; Forsberg, C.W. [Oak Ridge National Lab., Oak Ridge (United States)

    2004-07-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism.

  13. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    International Nuclear Information System (INIS)

    Il'kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I.; Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K.; Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A.; Haire, Jonathan M.; Forsberg, C.W.

    2004-01-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism

  14. Design of impact limiters of a bulk type B (U) . Trials of fall and validation of the analytical model In the design of a container for transportation of spent fuel, the impact limiters are a fundamental part for compliance with regulatory requirements; Diseno de los Limitadores de impacto de un Bulto Tipo B(U). Ensayos de Caida y validacion del Modelo Analitico

    Energy Technology Data Exchange (ETDEWEB)

    Garrido Quevedo, D.

    2013-07-01

    The aim is to confirm through real trials that the design and the results obtained through simulation conform to reality with a high degree of confidence... The combination of tests on scale models and the validation of the methods of calculation are necessary tools for the design of limiters impact a container of spent fuel transport.

  15. Decree no. 2001-1199 of the 10 december 2001 publishing the resolution MSC. 88 (71) notifying adoption of the international compilation of safety rules for the spent nuclear fuels, plutonium and high level radioactive wastes transport in casks on ships (compilation INF) (annexes), adopted at London the 27 may 1999

    International Nuclear Information System (INIS)

    2002-01-01

    This legislative text concerns the safety rules of spent nuclear fuels, plutonium and high level radioactive wastes transport, in casks on ships. Rules, fire prevention, temperature control of casks, electric supply, radioprotection, management and emergency plans are detailed. (A.L.B.)

  16. 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

  17. Spent fuel storage and isolation

    International Nuclear Information System (INIS)

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

    1979-02-01

    The principal spent fuel activities conducted within the commercial waste and spent fuel within the Commercial Waste and Spent Fuel Packaging Program are: simulated near-surface (drywell) storage demonstrations at Hanford and the Nevada Test Site; surface (sealed storage cask) and drywell demonstrations at the Nevada Test Site; and spent fuel receiving and packaging facility conceptual design. These investigations are described

  18. Cementation of residue ion exchange resins at Rocky Flats

    International Nuclear Information System (INIS)

    Dustin, D.F.; Beckman, T.D.; Madore, C.M.

    1998-01-01

    Ion exchange resins have been used to purify nitric acid solutions of plutonium at Rocky Flats since the 1950s. Spent ion exchange resins were retained for eventual recovery of residual plutonium, typically by incineration followed by the aqueous extraction of plutonium from the resultant ash. The elimination of incineration as a recovery process in the late 1980s and the absence of a suitable alternative process for plutonium recovery from resins led to a situation where spent ion exchange resins were simply placed into temporary storage. This report describes the method that Rocky Flats is currently using to stabilize residue ion exchange resins. The objective of the resin stabilization program is: (1) to ensure their safety during interim storage at the site, and (2) to prepare them for ultimate shipment to the Waste Isolation Pilot Plant (WIPP) in New Mexico. Included in the discussion is a description of the safety concerns associated with ion exchange resins, alternatives considered for their stabilization, the selection of the preferred treatment method, the means of implementing the preferred option, and the progress to date

  19. Fixing of various simulated radioactive wastes in urea-formaldehyde resin

    International Nuclear Information System (INIS)

    Du Dahai; Wei Peng

    1986-01-01

    This paper outlines the results of the fixing of a variety of simulated radioactive wastes in the urea-formaldehyde resin. The radioactive waste materials fixed include spent ion exchange resin, concentrates of NaNO 3 -NaBO 2 as well as NaBO 2 and sludge. The performance of the fixed products has been improved by means of selecting the synthetic conditions of resin, a suitable hardener and an inorganic additive

  20. Review: Resin Composite Filling

    Directory of Open Access Journals (Sweden)

    Desmond Ng

    2010-02-01

    Full Text Available The leading cause of oral pain and tooth loss is from caries and their treatment include restoration using amalgam, resin, porcelain and gold, endodontic therapy and extraction. Resin composite restorations have grown popular over the last half a century because it can take shades more similar to enamel. Here, we discuss the history and use of resin, comparison between amalgam and resin, clinical procedures involved and finishing and polishing techniques for resin restoration. Although resin composite has aesthetic advantages over amalgam, one of the major disadvantage include polymerization shrinkage and future research is needed on reaction kinetics and viscoelastic behaviour to minimize shrinkage stress.

  1. Review: Resin Composite Filling

    Science.gov (United States)

    Chan, Keith H. S.; Mai, Yanjie; Kim, Harry; Tong, Keith C. T.; Ng, Desmond; Hsiao, Jimmy C. M.

    2010-01-01

    The leading cause of oral pain and tooth loss is from caries and their treatment include restoration using amalgam, resin, porcelain and gold, endodontic therapy and extraction. Resin composite restorations have grown popular over the last half a century because it can take shades more similar to enamel. Here, we discuss the history and use of resin, comparison between amalgam and resin, clinical procedures involved and finishing and polishing techniques for resin restoration. Although resin composite has aesthetic advantages over amalgam, one of the major disadvantage include polymerization shrinkage and future research is needed on reaction kinetics and viscoelastic behaviour to minimize shrinkage stress.

  2. Spent Fuel in Chile

    International Nuclear Information System (INIS)

    López Lizana, F.

    2015-01-01

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

  3. Spent fuel pyroprocessing demonstration

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  4. Spent fuel reprocessing options

    International Nuclear Information System (INIS)

    2008-08-01

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

  5. Biodiesel Production from Spent Coffee Grounds

    Directory of Open Access Journals (Sweden)

    Blinová Lenka

    2017-06-01

    Full Text Available The residue after brewing the spent coffee grounds is an oil-containing waste material having a potential of being used as biodiesel feedstock. Biodiesel production from the waste coffee grounds oil involves collection and transportation of coffee residue, drying, oil extraction, and finally production of biodiesel. Different methods of oil extraction with organic solvents under different conditions show significant differences in the extraction yields. In the manufacturing of biodiesel from coffee oil, the level of reaction completion strongly depends on the quality of the feedstock oil. This paper presents an overview of oil extraction and a method of biodiesel production from spent coffee grounds.

  6. Biodiesel Production from Spent Coffee Grounds

    Science.gov (United States)

    Blinová, Lenka; Bartošová, Alica; Sirotiak, Maroš

    2017-06-01

    The residue after brewing the spent coffee grounds is an oil-containing waste material having a potential of being used as biodiesel feedstock. Biodiesel production from the waste coffee grounds oil involves collection and transportation of coffee residue, drying, oil extraction, and finally production of biodiesel. Different methods of oil extraction with organic solvents under different conditions show significant differences in the extraction yields. In the manufacturing of biodiesel from coffee oil, the level of reaction completion strongly depends on the quality of the feedstock oil. This paper presents an overview of oil extraction and a method of biodiesel production from spent coffee grounds.

  7. A STUDY ON ADSORPTION AND DESORPTION BEHAVIORS OF 14C FROM A MIXED BED RESIN

    Directory of Open Access Journals (Sweden)

    SEUNG-CHUL PARK

    2014-12-01

    Full Text Available Spent resin waste containing a high concentration of 14C radionuclide cannot be disposed of directly. A fundamental study on selective 14C stripping, especially from the IRN-150 mixed bed resin, was carried out. In single ion-exchange equilibrium isotherm experiments, the ion adsorption capacity of the fresh resin for non-radioactive HCO3− ion, as the chemical form of 14C, was evaluated as 11mg-C/g-resin. Adsorption affinity of anions to the resin was derived in order of NO3− > HCO3− ≥ H2PO4−. Thus the competitive adsorption affinity of NO3− ion in binary systems appeared far higher than that of HCO3− or H2PO4−, and the selective desorption of HCO3− from the resin was very effective. On one hand, the affinity of Co2+ and Cs+ for the resin remained relatively higher than that of other cations in the same stripping solution. Desorption of Cs+ was minimized when the summation of the metal ions in the spent resin and the other cations in solution was near saturation and the pH value was maintained above 4.5. Among the various solutions tested, from the view-point of the simple second waste process, NH4H2PO4 solution was preferable for the stripping of 14C from the spent resin.

  8. Sealed can of spent fuel

    International Nuclear Information System (INIS)

    Suzuki, Yasuyuki.

    1976-01-01

    Object: To provide a seal plug cover with a gripping portion fitted to a canning machine and a gripping portion fitted to a gripper of the same configuration as a fuel body for handling the fuel body so as to facilitate the handling work. Structure: A sealed can comprises a vessel and a seal plug cover, said cover being substantially in the form of a bottomed cylinder, which is slipped on the vessel and air-tightly secured by a fastening bolt between it and a flange. The spent fuel body is received into the vessel together with coolant during the step of canning operation. Said seal plug cover has two gripping portions, one for opening and closing the plug cover of the canning machine as an exclusive use member, the other being in the form of a hook-shaped peripheral groove, whereby the gripping portions may be effectively used using the same gripper when the spent fuel body is transported while being received in the sealed can or when the fuel body is removed from the sealed can. (Kawakami, Y.)

  9. Reprocessing of spent plasma

    International Nuclear Information System (INIS)

    Pierini, G.

    1981-01-01

    This invention relates to a process for removing helium and other impurities from a mixture containing deuterium and tritium, a deuterium/tritium mixture when purified in accordance with such a process and, more particularly, to a process for the reprocessing of spent plasma removed from a thermofusion reactor. (U.K.)

  10. Spent fuel counter

    International Nuclear Information System (INIS)

    Drayer, D.D.

    1988-09-01

    In many cases the IAEA must inspect spent fuel shipping casks before they leave facilities. Similarly, inspections may be required at the location where a cask is received and unloaded. In order to reduce the number of inspections required, it would be desirable to develop a system to count spent fuel assemblies as they are loaded or removed from shipping casks. This report discusses several methods which potentially could be used for performing this function. A concept for a Spent Fuel Counter System is proposed which uses a Laser Surveillance System (LASSY), Cerenkov Viewing Device (CVD), and Modular Integrated Video System (MIVS), all coupled together. In the proposed system, LASSY would provide an indication that an object is being placed into or removed from the cask, the CVD would be used to determine if the object has the radiation characteristics of a spent fuel assembly, and the MIVS would record the information. The system may need to be designed so that the operator could determine that it was operating correctly during the loading operations. This would help prevent anomalies from occurring which could only be resolved through reverification measures. Before such a system could be implemented testing would be necessary to determine that the individual components would each work adequately in this application. The issues of reliability, intrusiveness, and cask sealing should also be addressed before a development program is undertaken. 12 refs., 1 fig

  11. Time well spent

    DEFF Research Database (Denmark)

    Fallesen, Peter

    2013-01-01

    Individuals who spent time in foster care as children fare on average worse than non-placed peers in early adult life. Recent research on the effect of foster care placement on early adult life outcomes provides mixed evidence. Some studies suggest negative effects of foster care placement on early...

  12. Spent fuel storage at the Rancho Seco Nuclear Generation Station

    International Nuclear Information System (INIS)

    Miller, K.R.; Field, J.J.

    1995-01-01

    The Sacramento Municipal Utility District (SMUD) has developed a strategy for the storage and transport of spent nuclear fuel and is now in the process of licensing and manufacturing a Transportable Storage System (TSS). Staff has also engaged in impact limiter testing, non-fuel bearing component reinsertion, storage and disposal of GTCC waste, and site specific upgrades in support of spent fuel dry storage

  13. Feasibility and incentives for the consideration of spent fuel operating histories in the criticality analysis of spent fuel shipping casks

    International Nuclear Information System (INIS)

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

    1987-08-01

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

  14. Review of SR-Can: Evaluation of SKB's handling of spent fuel performance, radionuclide chemistry and geosphere transport parameters. External review contribution in support of SKI's and SSI's review of SR-Can

    International Nuclear Information System (INIS)

    Stenhouse, Mike; Jegou, Christophe; Brown, Paul; Meinrath, Guenther; Nitsche, Heino; Ekberg, Christian

    2008-03-01

    SR-Can covers the containment phase of the KBS-3 barriers as well as the consequences of releases of radionuclides to the rock and eventually the biosphere (after complete containment within fuel canisters has partially failed). The aim of this report is to provide a range of review comments with respect to those parameters related to spent fuel performance as well as radionuclide chemistry and transport. These parameter values are used in the quantification of consequences due to release of radionuclides from potentially leaking canisters. The report does not cover modelling approaches used for quantification of consequences. However, modelling used to derive parameter values is to some extent addressed (such as calculation of maximum radionuclide concentration due to formation of solubility limiting phases). The following are the key highlights and comments generated in the course of the review: Inconsistencies exist between recommendations provided in technical reports and those quoted in the Data Report. One of the reasons for such inconsistencies has been the timing of different pieces of research. It is hoped that the timing of contributions to SR-Site will be such that these inconsistencies can be avoided. Sensitivity analyses need to be carried out and reported in a number of areas to support some of the assumptions or decisions made in the assessment calculations. The likelihood is that SKB has performed many of the sensitivity analyses identified in different parts of this report, but these need to be reported, preferably to complement the recommendations provided

  15. Resin-Powder Dispenser

    Science.gov (United States)

    Standfield, Clarence E.

    1994-01-01

    Resin-powder dispenser used at NASA's Langley Research Center for processing of composite-material prepregs. Dispenser evenly distributes powder (resin polymer and other matrix materials in powder form) onto wet uncured prepregs. Provides versatility in distribution of solid resin in prepreg operation. Used wherever there is requirement for even, continuous distribution of small amount of powder.

  16. Studies on the Conditioning Methods of Spent Tri-butyl Phosphate/Kerosene and its Degradation Product in Different Matrices

    International Nuclear Information System (INIS)

    El-Dessouky, M.I.; El-sourougy, M.R.; Abed El-Aziz, M.M.; Aly, H.F.

    1999-01-01

    The destruction of spent TBP/Kerosene (odourless Kerosene (OK)) with potassium permanganate have been investigated. Comparative studies on the immobilization of spent TBP/Kerosene and its degradation product into different matrices have been carried out. The matrices used include, ordinary Portland cement, silica fume, treated fly ash, epoxy resin and cement mixed with epoxy resin.The different factors affecting solidified waste forms such as, compressive strength, water resistance, thermal stability, chemical resistance, radiological stability and leachability have been investigated. It was found that, epoxy resin and cement mixed with 5,10,20, and 50% of epoxy resin enhance the compressive strength of the solidified waste forms with spent TBP/OK more than that obtained from degradation products. The leaching rates of 152 and 154 Eu and 181 Hf from waste forms containing TBP/OK was found lower than that with degradation product

  17. Studies on the conditioning methods of spent tri-butyl phosphate/kerosene and its degradation product in different matrices

    International Nuclear Information System (INIS)

    El-Dessouky, M.I.; El-Sourougy, M.R.; Abed El-Aziz, M.M.; Aly, H.F.

    2001-01-01

    The destruction of spent TBP/Kerosene (odourless kerosene (OK)) with potassium permanganate has been investigated. Comparative studies on the immobilization of spent TBP/Kerosene and its degradation product into different matrices have been carried out. The matrices used include, ordinary portland cement, silica fume, treated fly ash, epoxy resin and cement mixed with epoxy resin. The different factors affecting solidified waste forms, such as compressive strength, water resistance, thermal stability, chemical resistance, radiological stability and leachability, have been investigated. It was found that epoxy resin and cement mixed with 5, 10, 20, and 50% of epoxy resin enhance the compressive strength of the solidified waste forms with spent TBP/OK more than that obtained from degradation products. The leaching rates of 152,154 Eu and 181 Hf from waste forms containing TBP/OK was found lower than that with degradation product. (author)

  18. Anion-exchange resin-based desulfurization process

    Energy Technology Data Exchange (ETDEWEB)

    Sheth, A.C.; Strevel, S.D.; Dharmapurikar, R.

    1992-01-01

    Under the current grant, the University of Tennessee Space Institute (UTSI) will carry out the bench scale evaluation and further development of the anion-exchange resin-based desulfurization concept to desulfurize alkali metal sulfates. This concept has been developed and patented by UTSI under US Patent No. 4,917,874. The developmental program proposed under this DOE grant includes screening of commercially available resins to select three candidate resins for further study. These three resins will undergo a series of experiments designed to test the resins' performance under different process conditions (including the use of spent MHD seed material). The best of these resins will be used in optimizing the regeneration step and in testing the effects of performance enhancers. The process schematic developed from the results will be used to estimate the related economics. During this reporting period, October 1, 1991 to September 30, 1992, analysis of batch mode screening experiments was completed to select three candidate resins for process variables study in the fixed-bed set-up. This setup was modified and the experiments were carded out to evaluate effects of major process variables. The analysis of fixed-bed experiments is going on and we have also started simple batch mode experiments to identify desirable conditions for resin regeneration step. We have also started simple process engineering type calculations to determine the trade-off between the solution concentration and the resulting evaporation/concentration load.

  19. Determination of degradation conditions of exchange resins containing technetium

    International Nuclear Information System (INIS)

    Rivera S, A.; Monroy G, F.; Quintero P, E.

    2014-10-01

    The quantification of Tc-99 in spent exchange resins, coming from nuclear power plants, is indispensable to define their administration. The Tc-99 is a pure beta emitter of 210000 years of half-life, volatile and of a high mobility in water and soil. For this reason, the objective of this work is to establish a digestion method of ionic exchange resins containing technetium that retains more than 95% of this radioisotope. Mineralization tests were carried out of a resin Amberlite IRN-150 by means of an oxidation heat, in acid medium, varying the resin mass, the medium volume, the media type, the temperature and the digestion time. The digested samples were analyzed by gas chromatography to estimate the grade of their degradation. The 99m Tc was used as tracer to determine the technetium percentage recovered after mineralizing the resin. The digestion process depends on the temperature and the resin mass. At higher temperature better mineralization of samples and to greater resin mass to a constant temperature, less degradation of the resin. The spectra beta of the 99m Tc and 99 Tc are presented. (Author)

  20. Encapsulating spent nuclear fuel

    International Nuclear Information System (INIS)

    Fleischer, L.R.; Gunasekaran, M.

    1979-01-01

    A system is described for encapsulating spent nuclear fuel discharged from nuclear reactors in the form of rods or multi-rod assemblies. The rods are completely and contiguously enclosed in concrete in which metallic fibres are incorporated to increase thermal conductivity and polymers to decrease fluid permeability. This technique provides the advantage of acceptable long-term stability for storage over the conventional underwater storage method. Examples are given of suitable concrete compositions. (UK)

  1. 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

  2. Biodiesel Production from Spent Coffee Grounds

    OpenAIRE

    Blinová Lenka; Bartošová Alica; Sirotiak Maroš

    2017-01-01

    The residue after brewing the spent coffee grounds is an oil-containing waste material having a potential of being used as biodiesel feedstock. Biodiesel production from the waste coffee grounds oil involves collection and transportation of coffee residue, drying, oil extraction, and finally production of biodiesel. Different methods of oil extraction with organic solvents under different conditions show significant differences in the extraction yields. In the manufacturing of biodiesel from ...

  3. Spent fuel packaging and its safety analysis

    International Nuclear Information System (INIS)

    Takada, Kimitaka; Nakaoki, Kozo; Tamamura, Tadao; Matsuda, Fumio; Fukudome, Kazuyuki

    1983-01-01

    An all stainless steel B(U) type packaging is proposed to transport spent fuels discharged from research reactors and other radioactive materials. The package is used dry and provided with surface fins to absorb drop shock and to dissipate decay heat. Safety was analyzed for structural, thermal, containment shielding and criticality factors, and the integrity of the package was confirmed with the MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, and KENO computer codes. (author)

  4. Overview of the United States spent nuclear fuel program

    International Nuclear Information System (INIS)

    Hurt, W.L.

    1997-12-01

    As a result of the end of the Cold War, the mission of the US Department of Energy (DOE) has shifted from an emphasis on nuclear weapons development and production to an emphasis on the safe management and disposal of excess nuclear materials including spent nuclear fuel from both production and research reactors. Within the US, there are two groups managing spent nuclear fuel. Commercial nuclear power plants are managing their spent nuclear fuel at the individual reactor sites until the planned repository is opened. All other spent nuclear fuel, including research reactors, university reactors, naval reactors, and legacy material from the Cold War is managed by DOE. DOE's mission is to safely and efficiently manage its spent nuclear fuel and prepare it for disposal. This mission involves correcting existing vulnerabilities in spent fuel storage; moving spent fuel from wet basins to dry storage; processing at-risk spent fuel; and preparing spent fuel in road-ready condition for repository disposal. Most of DOE's spent nuclear fuel is stored in underwater basins (wet storage). Many of these basins are outdated, and spent fuel is to be removed and transferred to more modern basins or to new dry storage facilities. In 1995, DOE completed a complex-wide environmental impact analysis that resulted in spent fuel being sent to one of three principal DOE sites for interim storage (up to 40 years) prior to shipment to a repository. This regionalization by fuel type will allow for economies of scale yet minimize unnecessary transportation. This paper discusses the national SNF program, ultimate disposition of SNF, and the technical challenges that have yet to be resolved, namely, release rate testing, non-destructive assay, alternative treatments, drying, and chemical reactivity

  5. Global Spent Fuel Logistics Systems Study (GSFLS). Volume 4. Pacific basin spent fuel logistics system

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-06-01

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

  6. Global Spent Fuel Logistics Systems Study (GSFLS). Volume 4. Pacific basin spent fuel logistics system

    International Nuclear Information System (INIS)

    1978-06-01

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

  7. Handling sticky resin by stingless bees (Hymenoptera, Apidae

    Directory of Open Access Journals (Sweden)

    Markus Gastauer

    2011-06-01

    Full Text Available For their nest defense, stingless bees (Meliponini collect plant resins which they stick on intruders like ants or cleptobiotic robber bees causing their immobilization. The aim of this article is to identify all parts of stingless bee workers contacting these sticky resins. Of special interest are those body parts with anti-adhesive properties to resin, where it can be removed without residues. For that, extensive behavioral observations during foraging flight, handling and application of the resin have been carried out. When handling the resin, all tarsi touch the resin while walking above it. For transportation from plants to the nest during foraging flight, the resin is packed to the corbicula via tarsi and basitarsi of front and middle legs. Once stuck to the resin or after the corbicula had been unloaded, the bee's legs have to be cleaned thoroughly. Only the tips of the mandibles, that form, cut and apply the sticky resin, seem to have at least temporarily resin-rejecting properties.

  8. The experiences from interim spent fuel storage operation with CASTOR 440/84 CASKS in NPP Dukovany

    International Nuclear Information System (INIS)

    Kuba, S.

    1999-01-01

    In this lecture are presented: principles of the CASTOR 440/84 design; design development works; commissioning of interim spent fuel storage facility; international transports of spent fuel utilising CASTOR 440/84 casks

  9. Case histories of West Valley spent fuel shipments: Final report

    International Nuclear Information System (INIS)

    1987-01-01

    In 1983, NRC/FC initiated a study on institutional issues related to spent fuel shipments originating at the former spent fuel processing facility in West Valley, New York. FC staff viewed the shipment campaigns as a one-time opportunity to document the institutional issues that may arise with a substantial increase in spent fuel shipping activity. NRC subsequently contracted with the Aerospace Corporation for the West Valley Study. This report contains a detailed description of the events which took place prior to and during the spent fuel shipments. The report also contains a discussion of the shipment issues that arose, and presents general findings. Most of the institutional issues discussed in the report do not fall under NRC's transportation authority. The case histories provide a reference to agencies and other institutions that may be involved in future spent fuel shipping campaigns. 130 refs., 7 figs., 19 tabs

  10. Case histories of West Valley spent fuel shipments: Final report

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    In 1983, NRC/FC initiated a study on institutional issues related to spent fuel shipments originating at the former spent fuel processing facility in West Valley, New York. FC staff viewed the shipment campaigns as a one-time opportunity to document the institutional issues that may arise with a substantial increase in spent fuel shipping activity. NRC subsequently contracted with the Aerospace Corporation for the West Valley Study. This report contains a detailed description of the events which took place prior to and during the spent fuel shipments. The report also contains a discussion of the shipment issues that arose, and presents general findings. Most of the institutional issues discussed in the report do not fall under NRC's transportation authority. The case histories provide a reference to agencies and other institutions that may be involved in future spent fuel shipping campaigns. 130 refs., 7 figs., 19 tabs.

  11. Flame retardant materials based on BDM/DBA resin and organic-inorganic additives

    OpenAIRE

    Wang, Zheng

    2017-01-01

    2,2’-diallyl bisphenol A (DBA) modified bismaleimide resins (BDM) are widely used as polymer matrix of high-performance composites in the field of aerospace, transportation, machinery, and electronics. This kind of resin has excellent properties, including good thermal stability, good processing performance, high mechanical strength, excellent dielectric properties, etc. However, as for most organic resins, the flame retardancy of the BMI resins is not so good. In order to expand the use of t...

  12. Spent fuel reprocessing method

    International Nuclear Information System (INIS)

    Shoji, Hirokazu; Mizuguchi, Koji; Kobayashi, Tsuguyuki.

    1996-01-01

    Spent oxide fuels containing oxides of uranium and transuranium elements are dismantled and sheared, then oxide fuels are reduced into metals of uranium and transuranium elements in a molten salt with or without mechanical removal of coatings. The reduced metals of uranium and transuranium elements and the molten salts are subjected to phase separation. From the metals of uranium and transuranium elements subjected to phase separation, uranium is separated to a solid cathode and transuranium elements are separated to a cadmium cathode by an electrolytic method. Molten salts deposited together with uranium to the solid cathode, and uranium and transuranium elements deposited to the cadmium cathode are distilled to remove deposited molten salts and cadmium. As a result, TRU oxides (solid) such as UO 2 , Pu 2 in spent fuels can be reduced to U and TRU by a high temperature metallurgical method not using an aqueous solution to separate them in the form of metal from other ingredients, and further, metal fuels can be obtained through an injection molding step depending on the purpose. (N.H.)

  13. Spent fuel interim storage

    International Nuclear Information System (INIS)

    Bilegan, Iosif C.

    2003-01-01

    The official inauguration of the spent fuel interim storage took place on Monday July 28, 2003 at Cernavoda NNP. The inaugural event was attended by local and central public authority representatives, a Canadian Government delegation as well as newsmen from local and central mass media and numerous specialists from Cernavoda NPP compound. Mr Andrei Grigorescu, State Secretary with the Economy and Commerce Ministry, underlined in his talk the importance of this objective for the continuous development of nuclear power in Romania as well as for Romania's complying with the EU practice in this field. Also the excellent collaboration between the Canadian contractor AECL and the Romanian partners Nuclear Montaj, CITON, UTI, General Concret in the accomplishment of this unit at the planned terms and costs. On behalf of Canadian delegation, spoke Minister Don Boudria. He underlined the importance which the Canadian Government affords to the cooperation with Romania aiming at specific objectives in the field of nuclear power such as the Cernavoda NPP Unit 2 and spent fuel interim storage. After traditional cutting of the inaugural ribbon by the two Ministers the festivities continued on the Cernavoda NPP Compound with undersigning the documents regarding the project completion and a press conference

  14. Mechanistic understanding of fouling of protein A chromatography resin.

    Science.gov (United States)

    Pathak, Mili; Rathore, Anurag S

    2016-08-12

    This paper aims to provide a thorough understanding of how fouling of Protein A resin takes place. Binding and mass transport properties of widely used agarose-based Protein A resin, MabSelect SuRe™, have been examined to understand the mechanism of resin fouling. There could be various factors that impact resin fouling. These include product/impurity build-up due to components in the feed material and ligand degradation due to the use of harsh buffers. To unravel their contributions, cycling studies were performed with and without product loading. The results presented in this paper provide a lucid understanding of the causative factors that limit Protein A chromatographic resin lifetime. The capacity fall for protein A resin at the end of 100th cycle due to use of feed material was found to be five times greater than that without using feed material. Compared to the fresh resin, the cycled resin samples shows 24% reduction in particle porosity and 51% reduction in pore mass transfer coefficient. Transmission electron microscopy (TEM) was used to qualitatively monitor accumulation of foulants on the cycled resin. Fouled resin sample contained a dense residue in the interior and exterior of resin particle both as a film at the bead surface and as granules. The surface activation energy increased five times in the case of fouled resin sample. The major event in fouling was identified as the non-specific adsorption of the feed material components on resin, signaling that pore diffusion is the rate limiting step. It is anticipated that these findings will assist in development of a more robust and economical downstream manufacturing process for monoclonal antibody purification. Copyright © 2016 Elsevier B.V. All rights reserved.

  15. Feasibility and kinetics study on the direct bio-regeneration of perchlorate laden anion-exchange resin.

    Science.gov (United States)

    Wang, Chao; Lippincott, Lee; Meng, Xiaoguang

    2008-11-01

    Anion exchange is one of the most promising treatment technologies for the removal of low levels of perchlorate. The spent anion-exchange resins, however, need to be disposed of or regenerated because they contain high contents of perchlorate. This study investigated the feasibility and kinetics of a direct bio-regeneration method. The method accomplished resin regeneration and biological perchlorate destruction concurrently, by directly contacting the spent resin with the perchlorate-reducing bacteria (PRB). The results indicated that the method was effective in regeneration of perchlorate and nitrate loaded resin and the resin could be repeatedly regenerated with the method. The regenerated resin was effective, stable, and durable in the filtration treatment of perchlorate in well water from the Saddle River area, NJ. Moreover, the method was also effective in regeneration of the spent A-530E resin, which had high perchlorate affinity and was yet very difficult for regeneration with the conventional brine desorption technique. Besides, the results further suggested that the perchlorate and nitrate desorption from the loaded resin coupling with their subsequent biological reduction could be the direct bio-regeneration mechanism. No biofilm was formed on the regenerated resin surface according to a scanning electron microscopy (SEM) analysis.

  16. REVIEW OF SPENT FUEL INTEGRITY EVALUATION FOR DRY STORAGE

    OpenAIRE

    DONGHAK KOOK; JONGWON CHOI; JUSEONG KIM; YONGSOO KIM

    2013-01-01

    Among the several options to solve PWR spent fuel accumulation problem in Korea, the dry storage method could be the most realistic and applicable solution in the near future. As the basic objectives of dry storage are to prevent a gross rupture of spent fuel during operation and to keep its retrievability until transportation, at the same time the importance of a spent fuel integrity evaluation that can estimate its condition at the final stage of dry storage is very high. According to the n...

  17. Comparison of spent fuel management fee collection alternatives

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  18. Handling, conditioning and disposal of spent sealed sources

    International Nuclear Information System (INIS)

    1990-02-01

    The series entitled ''Technical Manual for the Management of Low and Intermediate Level Wastes Generated at Small Nuclear Research Centres and by Radioisotope Users in Medicine, Research and Industry'' will serve as reference material to experts on technical assistance missions and provide ''direct know-how'' for technical staff in developing countries. This document is the first in the series. It provides the technical guidance and know-how necessary to permit developing Member States to safely handle, condition and store spent sealed radiation sources. It covers: characterization of sealed sources, legislation and regulations, management of spent sealed sources, transportation and disposal of spent sealed sources. 5 refs, 10 figs, 6 tabs

  19. Properties of light water reactor spent fuel cladding. Interim report

    International Nuclear Information System (INIS)

    Farwick, D.G.; Moen, R.A.

    1979-08-01

    The Commercial Waste and Spent Fuel Packaging Program will provide containment packages for the safe storage or disposal of spent Light Water Reactor (LWR) fuel. Maintaining containment of radionuclides during transportation, handling, processing and storage is essential, so the best understanding of the properties of the materials to be stored is necessary. This report provides data collection, assessment and recommendations for spent LWR fuel cladding materials properties. Major emphasis is placed on mechanical properties of the zircaloys and austenitic stainless steels. Limited information on elastic constants, physical properties, and anticipated corrosion behavior is also provided. Work is in progress to revise these evaluations as the program proceeds

  20. Foreign materials in a deep repository for spent nuclear fuels

    International Nuclear Information System (INIS)

    Jones, C.; Christiansson, Aa.; Wiborgh, M.

    1999-12-01

    The effects of foreign substances introduced into a spent-fuel repository are reviewed. Possible impacts on processes and barrier-functions are examined, and the following areas are identified: Corrosion of the spent-fuel canister through the presence of sulfur and substances that favor microbial growth; impacts on the bentonite properties through the presence of cations as calcium, potassium and iron; radionuclide transport through the presence of complex-formers and surface-active substances

  1. Quantitative Analysis of Radionuclide for the Used Resin of the Primary Purification System in HANARO

    International Nuclear Information System (INIS)

    Lee, Mun; Kim, Myong Seop; Park, Se Il; Kim, Tae Whan; Kim, Dong Hun; Kim, Young Chil

    2005-01-01

    In HANARO, a 30 MW research reactor, the ion exchange resin has been used for the purification of the primary coolant system. The resin used in the primary coolant purification system is replaced with new one once every 3 months during 30 MW reactor operation. The extracted resin from the primary coolant purification system is temporarily stored in a shielding treatment of the reactor hall for radiation cooling. After the radiation level of resin decreases enough to be handled for the waste disposal, it is put into the waste drum, and delivered to the waste facility in KAERI. Recently, in this procedure, the quantitative analysis of radionuclide which is contained in resin is required to have more quantitative data for the disposal. Therefore, in this work, a preliminary study was performed to find a sampling method for the representation of the characteristics of radionuclide in the spent resin

  2. Compacting spent fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    A method and apparatus for compacting spent fuel rods comprises transferring the rods from a nuclear fuel rod assembly into a different nuclear fuel rod container having a smaller cross section than the assembly. The individual rods are moved from a fuel assembly and through a transition funnel by movable grippers at opposite ends of the funnel. One movable gripper reciprocates between gripping and release positions in a gap between the fuel assembly and the transition funnel. All of the fuel rods are withdrawn concurrently and are merged towards one another into a tighter array within the transition funnel and emerge as a bundle. A movable and a stationary bundle gripper are provided between the funnel and the storage container to advance the bundle of fuel rods into the container. (author)

  3. Spent fuel storage at KURRI

    International Nuclear Information System (INIS)

    Nakagome, Y.; Fujita, Y.; Kanda, K.

    2004-01-01

    The Research Reactor Institute, Kyoto University (KURRI) has more than 200 MTR-type spent fuel elements stored in water pools. The longest pool residence time is 21 years at present. The integrity of spent fuel elements have been confirmed by a visual inspection and a sipping test. The spent fuel elements should be reprocessed in accordance with KURRI's policy. KURRI is now negotiating with a reprocessing plant to make a contract, as considering the consequences in U.S. (author)

  4. Numerical Estimation of the Spent Fuel Ratio

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilke, Jason [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Margraf, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunn, T. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-01-01

    Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank

  5. Survey of wet and dry spent fuel storage

    International Nuclear Information System (INIS)

    1999-07-01

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

  6. Preliminary Analysis of Radiation Shielding for B-type HIC Transport Package

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dohyung; Lee, Unjang; Ko, Jaehoon; Choi, Kyu-Sup [Korea Nuclear Engineering and Service Corporation, Seoul (Korea, Republic of)

    2007-10-15

    A radiation shielding analysis has been conducted using a computer program MCNP5 for a B-type HIC (High Integrated Container) Transport Package, which contains HIC with radioactive waste or Spent Resin, for transportation from nuclear power plant sites to disposal repository. Radiation source term is first carefully determined from the safety analysis reports related to HIC for appropriate calculation. And then MCNP (v.5) is performed to obtain the minimum thickness of HIC transport package, which meets the dose rate limit for HIC transport package prescribed in Korea Nuclear Law and IAEA Safety Standards for Radioactive Material Transport. In addition, some other analyses are done about the trend of dose rates depending on the thickness of shielding material and distance from the package.

  7. International symposium on storage of spent fuel from power reactors. Book of extended synopses

    International Nuclear Information System (INIS)

    1998-11-01

    This book of extended synopses includes papers presented at the International Symposium on Storage of Spent Fuel from Power Reactors organized by IAEA and held in Vienna from 9 to 13 November 1998. It deals with the problems of spent fuel management being an outstanding stage in the nuclear fuel cycle, strategy of interim spent fuel storage, transportation and encapsulation of spent fuel elements from power reactors. Spent fuel storage facilities at reactor sites are always wet while spent fuel storage facilities away from reactor are either wet or dry including casks and vaults. Different design solutions and constructions of storage or transportation casks as well as storing facilities are presented, as well as status of spent fuel storage together with experiences achieved in a number of member states, in the frame of safety, licensing and regulating procedures

  8. Historical overview of domestic spent fuel shipments

    International Nuclear Information System (INIS)

    Pope, R.B.; Wankerl, M.W.; Armstrong, S.; Hamberger, C.; Schmid, S.

    1991-01-01

    The purpose of this paper is to provide available historical data on most commercial and research reactor spent fuel shipments that have been completed in the United States between 1964 and 1989. This information includes data on the sources of spent fuel that has been shipped, the types of shipping casks used, the number of fuel assemblies that have been shipped, and the number of shipments that have been made. The data are updated periodically to keep abreast of changes. Information on shipments is provided for planning purposes; to support program decisions of the US Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM); and to inform interested members of the public, federal, state, and local government, Indian tribes, and the transportation community. 5 refs., 7 figs., 2 tabs

  9. Structural Evaluation on HIC Transport Packaging under Accident Conditions

    International Nuclear Information System (INIS)

    Chung, Sung Hwan; Kim, Duck Hoi; Jung, Jin Se; Yang, Ke Hyung; Lee, Heung Young

    2005-01-01

    HIC transport packaging to transport a high integrity container(HIC) containing dry spent resin generated from nuclear power plants is to comply with the regulatory requirements of Korea and IAEA for Type B packaging due to the high radioactivity of the content, and to maintain the structural integrity under normal and accident conditions. It must withstand 9 m free drop impact onto an unyielding surface and 1 m drop impact onto a mild steel bar in a position causing maximum damage. For the conceptual design of a cylindrical HIC transport package, three dimensional dynamic structural analysis to ensure that the integrity of the package is maintained under all credible loads for 9 m free drop and 1 m puncture conditions were carried out using ABAQUS code.

  10. Trunnions for spent fuel element shipping casks

    International Nuclear Information System (INIS)

    Cooke, B.

    1989-01-01

    Trunnions are used on spent fuel element shipping casks for one or more of a combination of lifting, tilting or securing to a transport vehicle. Within the nuclear transportation industry there are many different philosophies on trunnions, concerning the shape, manufacture, attachment, inspection, maintenance and repair. With the volume of international transport of spent fuel now taking place, it is recognized that problems are occurring with casks in international traffic due to the variance of the philosophies, national standards, and the lack of an international standard. It was agreed through the ISO that an international standard was required to harmonize. It was not possible to evolve an international standard. It was only possible to evolve an international guide. To evolve a standard would mean superseding any existing national standards which already cover particular aspects of trunnions i.e. deceleration forces imposed on trunnions used as tie down features. Therefore the document is a guide only and allows existing national standards to take precedence where they exist. The guide covers design, manufacture, maintenance, repair and quality assurance. The guide covers trunnions used on spent fuel casks transported by road, rail and sea. The guide details the considerations which should be taken account of by cask designers, i.e. stress intensity, design features, inspection and test methods etc. Manufacture, attachment and pre-service testing is also covered. The guide details user requirements which should also be taken account of, i.e. servicing frequency, content, maintenance and repair. The application of quality assurance is described separately although the principles are used throughout the guide

  11. The influence of the preparation of lignin bio-resins used in oil spilled agglomeration; A influencia do uso da lignina na preparacao de bioresinas utilizadas na aglomeracao de oleo derramado

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Joao Felipe T.; Silva, Thaissa P.; Ferreira, Leticia P.; Delazare, Thais; Souza Junior, Fernando G. [Instituto de Macromoleculas Professora Eloisa Mano - IMA/Universidade Federal do Rio de Janeiro - UFRJ, RJ (Brazil)], e-mail: fernando_gomes@ima.ufrj.br

    2011-07-01

    Locate petroleum, an energy source which spent millions of years to be formed, requires a lot of knowledge and technology, beyond large investments. However, along extraction, storage or transport of oil, there is a real risk of spills take place, causing extensive damage to the environment. Experience acquired due to accidental oil spills shows the large extension of damage imposed to the environment, impacting marine life, fishing and even tourism. Thus this study proposes the use of renewable sources, aiming to create an {sup g}reen absorbent material{sup .} This material is obtained through a polycondensation among lignin, furfural and cardanol, catalyzed by a strong acid. Synthesized bio-resin presents a good chemical similarity with oil, due to the tuning of its aromatic / aliphatic compounds, producing a strong physical interaction between the resin and oil, making the agglomeration process easy and contributing for the cleanup of oil spilled on water. (author)

  12. Survey of experience with dry storage of spent nuclear fuel and update of wet storage experience

    International Nuclear Information System (INIS)

    1988-01-01

    Spent fuel storage is an important part of spent fuel management. At present about 45,000 t of spent water reactor fuel have been discharged worldwide. Only a small fraction of this fuel (approximately 7%) has been reprocessed. The amount of spent fuel arisings will increase significantly in the next 15 years. Estimates indicate that up to the year 2000 about 200,000 t HM of spent fuel could be accumulated. In view of the large quantities of spent fuel discharged from nuclear power plants and future expected discharges, many countries are involved in the construction of facilities for the storage of spent fuel and in the development of effective methods for spent fuel surveillance and monitoring to ensure that reliable and safe operation of storage facilities is achievable until the time when the final disposal of spent fuel or high level wastes is feasible. The first demonstrations of final disposal are not expected before the years 2000-2020. This is why the long term storage of spent fuel and HLW is a vital problem for all countries with nuclear power programmes. The present survey contains data on dry storage and recent information on wet storage, transportation, rod consolidation, etc. The main aim is to provide spent fuel management policy making organizations, designers, scientists and spent fuel storage facility operators with the latest information on spent fuel storage technology under dry and wet conditions and on innovations in this field. Refs, figs and tabs

  13. Reduction of polyester resin shrinkage by means of epoxy resin

    International Nuclear Information System (INIS)

    Pietrzak, M.; Brzostowski, A.

    1981-01-01

    An attempt was made to decrease the shrinkage of unsaturated polyester resin, taking place during radiation-induced curing, by the addition of epoxy resin. In order to combine chemically both resins, the epoxy component was modified with cinnamic and acrylic acids. A composition of 90 parts of polyester resin, 10 parts of epoxy resin modified with cinnamic acid, and 150 parts of a silica filler showed a volume shrinkage of 1.2%. (author)

  14. WWER spent fuel storage

    International Nuclear Information System (INIS)

    Bower, C.C.; Lettington, C.

    1994-01-01

    Selection criteria for PAKS NPP dry storage system are outlined. They include the following: fuel temperature in storage; sub-criticality assurance (avoidance of criticality for fuel in the unirradiated condition without having to take credit for burn-up); assurance of decay heat removal; dose uptake to the operators and public; protection of environment; volume of waste produced during operation and decommissioning; physical protection of stored irradiated fuel assemblies; IAEA safeguards assurance; storage system versus final disposal route; cost of construction and extent of technology transfer to Hungarian industry. Several available systems are evaluated against these criteria, and as a result the GEC ALSTHOM Modular Vault Dry Store (MVDS) system has been selected. The MVDS is a passively cooled dry storage facility. Its most important technical, safety, licensing and technology transfer characteristics are outlined. On the basis of the experience gained some key questions and considerations related to the East European perspective in the field of spent fuel storage are discussed. 8 figs

  15. Transportation

    International Nuclear Information System (INIS)

    Anon.

    1998-01-01

    Here is the decree of the thirtieth of July 1998 relative to road transportation, to trade and brokerage of wastes. It requires to firms which carry out a road transportation as well as to traders and to brokers of wastes to declare their operations to the prefect. The declaration has to be renewed every five years. (O.M.)

  16. Containing method for spent fuel and spent fuel containing vessel

    International Nuclear Information System (INIS)

    Maekawa, Hiromichi; Hanada, Yoshine.

    1996-01-01

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

  17. HIGH ASPECT RATIO ION EXCHANGE RESIN BED - HYDRAULIC RESULTS FOR SPERICAL RESIN BEADS

    International Nuclear Information System (INIS)

    Duignan, M; Charles Nash, C; Timothy Punch, T

    2007-01-01

    A principal role of the DOE Savannah River Site is to safely dispose of a large volume of liquid nuclear waste held in many storage tanks. An in-tank ion exchange unit is being considered for cesium removal to accelerate waste processing. This unit is planned to have a relatively high bed height to diameter ratio (10:1). Complicating the design is the need to cool the ion exchange media; therefore, the ion exchange column will have a central cooling core making the flow path annular. To separate cesium from waste the media being considered is made of resorcinol formaldehyde resin deposited on spherical plastic beads and is a substitute for a previously tested resin made of crystalline silicotitanate. This spherical media not only has an advantage of being mechanically robust, but, unlike its predecessor, it is also reusable, that is, loaded cesium can be removed through elution and regeneration. Resin regeneration leads to more efficient operation and less spent resin waste, but its hydraulic performance in the planned ion exchange column was unknown. Moreover, the recycling process of this spherical resorcinol formaldehyde causes its volume to significantly shrink and swell. To determine the spherical media's hydraulic demand a linearly scaled column was designed and tested. The waste simulant used was prototypic of the wastes' viscosity and density. This paper discusses the hydraulic performance of the media that will be used to assist in the design of a full-scale unit

  18. Air Shipment of Spent Nuclear Fuel from Romania to Russia

    Energy Technology Data Exchange (ETDEWEB)

    Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

    2010-10-01

    Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities for shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.

  19. Present status of JMTR spent fuel shipment

    International Nuclear Information System (INIS)

    Miyazawa, Masataka; Watanabe, Masao; Yokokawa, Makoto; Sato, Hiroshi; Ito, Haruhiko

    2002-01-01

    The Japan Atomic Energy Research Institute (JAERI) has been consistently making the enrichment reduction of reactor fuels in cooperation with RERTR Program and FRR SNF Acceptance Program both conducted along with the U.S. Nuclear Non-Proliferation Policy and JMTR, 50 MW test reactor in Oarai Research Establishment, has achieved core conversion, from its initial 93% enriched UAl alloy to 45% enriched uranium-aluminide fuel, and then to the current 19.8% enriched uranium-silicide fuel. In order to return all of JMTR spent fuels, to be discharged from the reactor by May 12, 2006, to the U.S.A. by May 12, 2009, JAERI is planning the transportation schedule based on one shipment per year. The sixth shipment of spent fuels to U.S. was carried out as scheduled this year, where the total number of fuels shipped amounts to 651 elements. All of the UAl alloy elements have so far been shipped and now shipments of 45% enriched uranium-aluminide type fuels are in progress. Thus far the JMTR SFs have been transported on schedule. From 2003 onward are scheduled more then 850 elements to be shipped. In this paper, we describe our activities on the transportation in general and the schedule for the SFs shipments. (author)

  20. Resin composite repair: Quantitative microleakage evaluation of resin-resin and resin-tooth interfaces with different surface treatments

    OpenAIRE

    Celik, Cigdem; Cehreli, Sevi Burcak; Arhun, Neslihan

    2015-01-01

    Objective: The aim was to evaluate the effect of different adhesive systems and surface treatments on the integrity of resin-resin and resin-tooth interfaces after partial removal of preexisting resin composites using quantitative image analysis for microleakage testing protocol. Materials and Methods: A total of 80 human molar teeth were restored with either of the resin composites (Filtek Z250/GrandioSO) occlusally. The teeth were thermocycled (1000?). Mesial and distal 1/3 parts of the res...

  1. Recovery of decommissioning and spent fuel charges

    International Nuclear Information System (INIS)

    Bermanis, G.

    1982-01-01

    The licensing and financial aspects of NPP decommissioning, deactivation and dismantling of radioactive equipment in the USA are considered. Data on the costs of spent fuel transport and conservation are given. The state of the problem development in other countries is briefly described. It is pointed out that the technical aspect of the problem is much better studied than that of license-financial problem. At the same time in contrast to TPP NPP use is connected with considerable expenses after the end of a power plant sevice time

  2. 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

  3. Disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Transportation

    National Research Council Canada - National Science Library

    Allshouse, Michael; Armstrong, Frederick Henry; Burns, Stephen; Courts, Michael; Denn, Douglas; Fortunato, Paul; Gettings, Daniel; Hansen, David; Hoffman, Douglas; Jones, Robert

    2007-01-01

    .... The ability of the global transportation industry to rapidly move passengers and products from one corner of the globe to another continues to amaze even those wise to the dynamics of such operations...

  5. 75 FR 18242 - Florida Power and Light, St. Lucie Units 1 and 2; Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2010-04-09

    ...)(2)(i)(A), 72.212(b)(7), and 72.214, and would allow them to load spent fuel into the TN NUHOMS... and Light, St. Lucie Units 1 and 2; Independent Spent Fuel Storage Installation; Environmental... Davis, Senior Project Manager, Division of Spent Fuel Storage and Transportation, Office of Nuclear...

  6. Data requirements and maintenance of records for spent fuel management: A review

    International Nuclear Information System (INIS)

    2006-11-01

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

  7. ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  8. ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  9. Optimization of the resin application on the purification system of ISFSF RSG-GAS

    International Nuclear Information System (INIS)

    Junaidi

    1998-01-01

    The purification system in the spent fuel storage is to maintain the coolant water quality to the constant. This system uses the resin of IRN 77-Anion in the mixed bed filter as contamination capture. The returned water flow to the storage will be analyzed to the monitor the bed efficiency. This system can be optimally operated if the mixed bed is contained by 200 liters of resin and this is equivalent with the 0.5 m high of bed and 0.75 m in diameter. The resin must be flushed to be cleaned or changed when the pressure difference in the bed filter is higher than 0.5 bar

  10. Spent fuel management in Spain

    International Nuclear Information System (INIS)

    Gonzalez, J.L.

    2002-01-01

    The spent fuel management strategy in Spain is presented. The strategy includes temporary solutions and plans for final disposal. The need for R and D including partitioning and transmutation, as well as the financial constraints are also addressed. (author)

  11. IAEA spent fuel storage glossary

    International Nuclear Information System (INIS)

    1985-10-01

    The aim of this glossary is to provide a basis for improved international understanding of terms used in the important area of spent fuel storage technology. The glossary is the product of an IAEA Consultant Group with valuable input from a substantial list of reviewers. The glossary emphasizes fuel storage relevant to power reactors, but is also widely applicable to research reactors. The intention is to define terms from current technologies. Terms are limited to those directly related to spent fuel storage

  12. Decree no. 2001-1199 of the 10 december 2001 publishing the resolution MSC. 88 (71) notifying adoption of the international compilation of safety rules for the spent nuclear fuels, plutonium and high level radioactive wastes transport in casks on ships (compilation INF) (annexes), adopted at London the 27 may 1999; Decret no. 2001-1199 du 10 decembre 2001 portant publication de la resolution MSC.88 (71) portant adoption du recueil international de regles de securite pour le transport de combustible nucleaire irradie, de plutonium et de dechets hautement radioactifs en colis a bord de navires (recueil INF) (ensemble une annexe), adoptee a Londres le 27 mai 1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    This legislative text concerns the safety rules of spent nuclear fuels, plutonium and high level radioactive wastes transport, in casks on ships. Rules, fire prevention, temperature control of casks, electric supply, radioprotection, management and emergency plans are detailed. (A.L.B.)

  13. Characterization of radioactive contaminants and water treatment trials for the Taiwan Research Reactor's spent fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Chun-Ping, E-mail: chunping@iner.gov.tw [Institute of Nuclear Energy Research, 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC (China); Lin, Tzung-Yi; Chiao, Ling-Huan; Chen, Hong-Bin [Institute of Nuclear Energy Research, 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC (China)

    2012-09-30

    Highlights: Black-Right-Pointing-Pointer Deal with a practical radioactive contamination in Taiwan Research Reactor spent fuel pool water. Black-Right-Pointing-Pointer Identify the properties of radioactive contaminants and performance test for water treatment materials. Black-Right-Pointing-Pointer The radioactive solids were primary attributed by ruptured spent fuels, spent resins, and metal debris. Black-Right-Pointing-Pointer The radioactive ions were major composed by uranium and fission products. Black-Right-Pointing-Pointer Diatomite-based ceramic depth filter can simultaneously removal radioactive solids and ions. - Abstract: There were approximately 926 m{sup 3} of water contaminated by fission products and actinides in the Taiwan Research Reactor's spent fuel pool (TRR SFP). The solid and ionic contaminants were thoroughly characterized using radiochemical analyses, scanning electron microscopy equipped with an energy dispersive spectrometer (SEM-EDS), and inductively coupled plasma optical emission spectrometry (ICP-OES) in this study. The sludge was made up of agglomerates contaminated by spent fuel particles. Suspended solids from spent ion-exchange resins interfered with the clarity of the water. In addition, the ionic radionuclides such as {sup 137}Cs, {sup 90}Sr, U, and {alpha}-emitters, present in the water were measured. Various filters and cation-exchange resins were employed for water treatment trials, and the results indicated that the solid and ionic contaminants could be effectively removed through the use of <0.9 {mu}m filters and cation exchange resins, respectively. Interestingly, the removal of U was obviously efficient by cation exchange resin, and the ceramic depth filter composed of diatomite exhibited the properties of both filtration and adsorption. It was found that the ceramic depth filter could adsorb {beta}-emitters, {alpha}-emitters, and uranium ions. The diatomite-based ceramic depth filter was able to simultaneously

  14. Behavior study of spend ion exchange resins immobilized in pyrolyzed polymer matrix

    International Nuclear Information System (INIS)

    Ramos, P.B; Fuentes, N.O; Luca, V.

    2012-01-01

    The pyrolysis of spent ion exchange resins contained in epoxy resins represents an attractive alternative to cementation as a confining method. In this sense, a significant reduction of volume can be achieved, as well as avoiding the dispersion of the exhausted ion exchange resin by the means of an epoxy resin used as a matrix, while potentially limiting the release of highly radioactive long life isotopes such us Cs-137, Sr-90 and Co-60 among others. Three types of monoliths were made: (i) epoxy resin, (ii) epoxy resin with carbon and (iii) a binder of epoxy resin and clay. In every case, the monolith contained the ion exchange resin. They were prepared by the mixing of resin pearl loaded with epoxy cations and a subsequent pyrolysis process with a temperature increase ratio of 2 o C /min reaching maximum values in the range between 200 o C - 800 o C, remaining in it for 1 hour. Monoliths obtained for each final temperature had been characterized to obtain data corresponding to the mass loss, volume reduction and lixiviation, as well as mechanical and microstructural properties (author)

  15. Resin secretory structures of Boswellia papyrifera and implications for frankincense yield.

    Science.gov (United States)

    Tolera, Motuma; Menger, David; Sass-Klaassen, Ute; Sterck, Frank J; Copini, Paul; Bongers, Frans

    2013-01-01

    Frankincense, a gum-resin, has been tapped from Boswellia papyrifera trees for centuries. Despite the intensive tapping and economic interest of B. papyrifera, information on the resin secretory structures, which are responsible for synthesis, storage and transport of frankincense, is virtually absent. This study describes the type, architecture and distribution of resin secretory structures of B. papyrifera and its relevance for the ecophysiology and economic use of the tree. The type and architecture of resin secretory structures present in bark and wood was investigated from transversal, tangential and radial sections of bark and wood samples. The diameter and density (number of resin canals mm(-2)) of axial resin canals were determined from digital images of thin sections across the different zones of inner bark. Resin canals form a three-dimensional network within the inner bark. Yet, the intact resin-conducting and producing network is on average limited to the inner 6·6 mm of the inner bark. Within the inner bark, the density of non-lignified axial resin canals decreases and the density of lignified resin canals increases from the vascular cambium towards the outer bark. In the wood, only radial resin canals were encountered. Frankincense tapping techniques can be improved based on knowledge of bark anatomy and distribution and architecture of resin secretory structures. The suggested new techniques will contribute to a more sustainable frankincense production that enhances the contribution of frankincense to rural livelihoods and the national economy.

  16. Transportation of vitrified residues towards Switzerland

    International Nuclear Information System (INIS)

    2002-01-01

    This document presents the swiss policy concerning the spent fuel management and the participation of COGEMA to the spent fuel processing. In this framework the swiss spent fuel processing at COGEMA La Hague, the transports safety and management, the regulatory framework applied to the transports, the quality and safety control and the ultimate residues storage are presented. (A.L.B.)

  17. Transport of radioactive substances; Der Transport radioaktiver Stoffe

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-12-15

    The report on the transport of radioactive substances covers the following topics: facts on radioactive materials transport, safety of the transport of radioactive substances, legal regulations and guidelines: a multiform but consistent system, transport of nuclear fuels, safety during the transport of nuclear fuel, future transport of spent fuel elements and high-level radioactive wastes in Germany.

  18. Evaluation of ferrocyanide anion exchange resins regarding the uptake of Cs{sup +} ions and their regeneration

    Energy Technology Data Exchange (ETDEWEB)

    Won, Hui Jun; Mooon, Jei Kwon; Jung, Chong Hun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chung, Won Yang [Kangwon University, Chuncheon (Korea, Republic of)

    2008-10-15

    Ferrocyanide-anion exchange resin was prepared and the prepared ion exchange resins were tested on the ability to uptake Cs{sup +} ion. The prepared ion exchange resins were resin-KCoFC, resin-KNiFC, and resin-KCuFC. The three tested ion exchange resins showed ion exchange selectivity on the Cs{sup +} ion of the surrogate soil decontamination solution, and resin- KCoFC showed the best Cs{sup +} ion uptake capability among the tested ion exchange resins. The ion exchange behaviors were explained well by the modified Dubinin-Polanyi equation. A regeneration feasibility study of the spent ion exchange resins was also performed by the successive application of hydrogen peroxide and hydrazine. The desorption of the Cs{sup +} ion from the ion exchange resin satisfied the electroneutrality condition in the oxidation step; the desorption of the Fe{sup 2+} ion in the reduction step could also be reduced by adding the K{sup +} ion.

  19. Preliminary concepts for detecting national diversion of LWR spent fuel

    International Nuclear Information System (INIS)

    Sonnier, C.S.; Cravens, M.N.

    1978-04-01

    Preliminary concepts for detecting national diversion of LWR spent fuel during storage, handling and transportation are presented. Principal emphasis is placed on means to achieve timely detection by an international authority. This work was sponsored by the Department of Energy/Office of Safeguards and Security (DOE/OSS) as part of the overall Sandia Fixed Facility Physical Protection Program

  20. Commercial solutions [for dry spent fuel storage casks

    International Nuclear Information System (INIS)

    Howe, W.F.; Pennington, C.W.; Hobbs, J.; Lee, W.; Thomas, B.D.; Dibert, D.J.

    1996-01-01

    In the aftermath of the termination of the DOE's MPC (Multi-Purpose Canister) programme, commercial suppliers are coming forward with new or updated systems to meet utility needs. Leading vendors describe the advantages of their systems for dry spent fuel storage and transport. (Author)

  1. Method for removing contaminants from plastic resin

    Science.gov (United States)

    Bohnert, George W [Harrisonville, MO; Hand, Thomas E [Lee's Summit, MO; DeLaurentiis, Gary M [Jamestown, CA

    2008-12-09

    A resin recycling method that produces essentially contaminant-free synthetic resin material in an environmentally safe and economical manner. The method includes receiving the resin in container form. The containers are then ground into resin particles. The particles are exposed to a solvent, the solvent contacting the resin particles and substantially removing contaminants on the resin particles. After separating the particles and the resin, a solvent removing agent is used to remove any residual solvent remaining on the resin particles after separation.

  2. Application of Spatial Data Modeling Systems, Geographical Information Systems (GIS), and Transportation Routing Optimization Methods for Evaluating Integrated Deployment of Interim Spent Fuel Storage Installations and Advanced Nuclear Plants

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Gary T [ORNL; Belles, Randy [ORNL; Cetiner, Sacit M [ORNL; Howard, Rob L [ORNL; Liu, Cheng [ORNL; Mueller, Don [ORNL; Omitaomu, Olufemi A [ORNL; Peterson, Steven K [ORNL; Scaglione, John M [ORNL

    2012-06-01

    The objective of this siting study work is to support DOE in evaluating integrated advanced nuclear plant and ISFSI deployment options in the future. This study looks at several nuclear power plant growth scenarios that consider the locations of existing and planned commercial nuclear power plants integrated with the establishment of consolidated interim spent fuel storage installations (ISFSIs). This research project is aimed at providing methodologies, information, and insights that inform the process for determining and optimizing candidate areas for new advanced nuclear power generation plants and consolidated ISFSIs to meet projected US electric power demands for the future.

  3. Cost effectiveness of detritiating water with resin columns

    International Nuclear Information System (INIS)

    Drake, R.H.; Williams, D.S.

    1997-10-01

    There are technologies in use for cleaning up concentrated tritiated process water. These are not cost effective for tritiated water with low concentrations of tritium. There are currently no cost-effective technologies for cleaning up low-tritium-concentration tritiated water, such as most tritiated groundwater, spent fuel storage basin water, or underground storage tank water. Resin removal of tritium from tritiated water at low concentrations (near the order of magnitude of drinking water standard maximums) is being tested on TA-SO (Los Alamos National Laboratory's Liquid Radioactive Waste Treatment Facility) waste streams. There are good theoretical and test indications that this may be a technologically effective means of removing tritium from tritiated water. Because of likely engineering design similarity, it is reasonable to anticipate that a resin column system's costs will be similar to some common commercial water treatment systems. Thus, the potential cost effectiveness of a resin treatment system offers hope for treating tritiated water at affordable costs. The TA-50 resin treatment cost projection of $18 per 1,000 gallons is within the same order of magnitude as cost data for typical commercial groundwater cleanup projects. The prospective Los Alamos National Laboratory (LANL) resin treatment system at $18 per 1,000 gallons appears to have a likely cost advantage of at least an order of magnitude over the competing, developmental, water detritiation technologies

  4. Container for spent fuel assembly

    International Nuclear Information System (INIS)

    Sawai, Takeshi.

    1996-01-01

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

  5. Safety and economic study of special trains for shipment of spent fuel

    International Nuclear Information System (INIS)

    Loscutoff, W.V.; Murphy, E.S.; Clark, L.L.; McKee, R.W.; Hall, R.J.

    1977-12-01

    Spent fuel from nuclear reactors has been transported in routine commerce in the U.S. for over 25 y. During this period, no accident has occurred that resulted in the release of radioactive material from a spent fuel cask. By 1986, the annual rate of spent fuel shipments by rail is expected to be nearly 20 percent of the total shipments during the last 25 y. Because of this increased shipping rate, the railroad industry contends that the transportation environment for spent fuel shipments must be controlled to maintain this excellent safety record. To provide this control over the transportation environment, the Association of American Railroads (AAR) has proposed that all spent fuel be handled in special train service. A set of operating practices for spent fuel special trains has been proposed that includes a 35 mph maximum speed restriction, a passing restriction and a no-other-freight restriction. The purpose of this study was to evaluate the potential effects of these restrictions on the safety and cost of shipping spent fuel by train. The analysis is based on estimated quantities of spent fuel transported in 1986

  6. Spent Nuclear Fuel project, project management plan

    International Nuclear Information System (INIS)

    Fuquay, B.J.

    1995-01-01

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

  7. Contact allergy to epoxy resin

    DEFF Research Database (Denmark)

    Bangsgaard, Nannie; Thyssen, Jacob Pontoppidan; Menné, Torkil

    2012-01-01

    to epoxy resin remained stable over the study period. Of the patients with an epoxy resin-positive patch test, 71% returned a questionnaire; 95 patients had worked with epoxy resin in the occupational setting, and, of these, one-third did not use protective gloves and only 50.5% (48) had participated...

  8. Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio programme

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  9. Spent Pot Lining Characterization Framework

    Science.gov (United States)

    Ospina, Gustavo; Hassan, Mohamed I.

    2017-09-01

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

  10. Spent fuel management in Japan

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  11. New strategies for treatment and reuse of spent sulfidic caustic stream from petroleum industry

    Directory of Open Access Journals (Sweden)

    Jéssica Frontino Paulino

    2012-01-01

    Full Text Available This work examines traditional and new routes for removal of H2S and other sulfur compounds from spent sufidic caustic (SSC. SH- (hydrogenosulfide and S2- (sulfide ions were quantitatively oxidized at 25 ºC using H2O2, NaOCl or a spent sulfochromic mixture. SH-/S2- ions were also removed via reaction with freshly prepared iron or manganese hydroxides, or after passing the SSC through strong basic anion exchange resins (OH- form. The treated caustic solution, as well as iron/manganese hydroxides, removed H2S from diesel samples at 25 ºC. SSC treatment via strong basic anion-exchange resins produced the treated caustic solution with the highest free alkalinity.

  12. Enhancement of adhesion between resin coating materials and resin cements.

    Science.gov (United States)

    Udo, Tomoaki; Nikaido, Toru; Ikeda, Masaomi; Weerasinghe, Dinesh S; Harada, Naoko; Foxton, Richard M; Tagami, Junji

    2007-07-01

    Resin coating technique is a unique method that improves the dentin bond strength of resin cements in indirect restorations. However, the weak link of a specimen bonded using the resin coating technique was reported to be the bonded interface between the resin coating material and resin cement. The purpose of this study, therefore, was to enhance the bonding performance between a resin coating material and a resin cement. Two light-cured flowable composites, Protect Liner F and Clearfil Flow FX, were used as coating materials, and two dual-cure composite materials, Panavia F 2.0 and Clearfil DC Core Automix, were used as resin cements. The ultimate tensile strength of each material and the microtensile bond strengths of the bonded specimens of resin coating material and resin cement were measured using a crosshead speed of 1.0 mm/min. Three-way ANOVA (p=0.05) revealed that the highest microtensile bond strength was obtained using a combination of Clearfil Flow FX and Clearfil DC Core Automix, and when the surface of the coating material was treated with ED Primer II. It was strongly suggested that materials with a higher ultimate tensile strength, when used in both resin coating and cementation, could enhance the bond strength between the two.

  13. Resin impregnation process for producing a resin-fiber composite

    Science.gov (United States)

    Palmer, Raymond J. (Inventor); Moore, William E. (Inventor)

    1994-01-01

    Process for vacuum impregnation of a dry fiber reinforcement with a curable resin to produce a resin-fiber composite, by drawing a vacuum to permit flow of curable liquid resin into and through a fiber reinforcement to impregnate same and curing the resin-impregnated fiber reinforcement at a sufficient temperature and pressure to effect final curing. Both vacuum and positive pressure, e.g. autoclave pressure, are applied to the dry fiber reinforcement prior to application of heat and prior to any resin flow to compact the dry fiber reinforcement, and produce a resin-fiber composite of reduced weight, thickness and resin content, and improved mechanical properties. Preferably both a vacuum and positive pressure, e.g. autoclave pressure, are also applied during final curing.

  14. Dry transfer system for spent fuel: Project report, A system designed to achieve the dry transfer of bare spent fuel between two casks. Final report

    International Nuclear Information System (INIS)

    Dawson, D.M.; Guerra, G.; Neider, T.; Shih, P.

    1995-12-01

    This report describes the system developed by EPRI/DOE for the dry transfer of spent fuel assemblies outside the reactor spent fuel pool. The system is designed to allow spent fuel assemblies to be removed from a spent fuel pool in a small cask, transported to the transfer facility, and transferred to a larger cask, either for off-site transportation or on-site storage. With design modifications, this design is capable of transferring single spent fuel assemblies from dry storage casks to transportation casks or visa versa. One incentive for the development of this design is that utilities with limited lifting capacity or other physical or regulatory constraints are limited in their ability to utilize the current, more efficient transportation and storage cask designs. In addition, DOE, in planning to develop and implement the multi-purpose canister (MPC) system for the Civilian Radioactive Waste Management System, included the concept of an on-site dry transfer system to support the implementation of the MPC system at reactors with limitations that preclude the handling of the MPC system transfer casks. This Dry Transfer System can also be used at reactors wi decommissioned spent fuel pools and fuel in dry storage in non-MPC systems to transfer fuel into transportation casks. It can also be used at off-reactor site interim storage facilities for the same purpose

  15. Nuclear transport

    International Nuclear Information System (INIS)

    Anon.

    2001-01-01

    Here is given the decree (2001/1199) of the 10. of december 2001 relative to the passing of safety rules concerning the maritime transport of spent fuels, plutonium and high-level radioactive wastes contained in packages. (O.M.)

  16. Spent fuel management in Spain

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. The safety of storage facilities for LMFBR spent fuels

    International Nuclear Information System (INIS)

    Lefort, G.; Puit, J.C.

    1982-04-01

    Storage conditions for the cooling of fast neutron spent fuel assembly, before reprocessing, will be adapted specifically to this fuel cycle, but in a preliminary testing period it is better to take advantage of the industrial storage conditions already used for spent fuel of light water reactors, even if no extrapolation is possible in the future. In this aim fuel assemblies are dismantled, as soon as thermal conditions allow it, and pins are gathered in leak proof containers in a gas atmosphere and put in a cooling pool. This solution gives good results but studies and experiments are resumed, for future processing, for safe, easy and economical transport and storage conditions [fr

  18. Transportation of vitrified residues towards Switzerland; Le transport des residus vitrifies vers la Suisse

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    This document presents the swiss policy concerning the spent fuel management and the participation of COGEMA to the spent fuel processing. In this framework the swiss spent fuel processing at COGEMA La Hague, the transports safety and management, the regulatory framework applied to the transports, the quality and safety control and the ultimate residues storage are presented. (A.L.B.)

  19. Technical strategy for the management of INEEL spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    This report presents evaluations, findings, and recommendations of the Idaho National Engineering and Environmental Laboratory (INEEL) Spent Nuclear Fuel Task Team. The technical strategy developed by the Task Team includes stabilization, near term storage, packaging, transport, and ultimate disposal. Key issues identified and discussed include waste characterization, criticality, packaging, waste form performance, and special fuels. Current plans focus on onsite needs, and include three central elements: (1) resolution of near-term vulnerabilities, (2) consolidation of storage locations, and (3) achieving dry storage in transportable packages. In addition to the Task Team report, appendices contain information on the INEEL spent fuel inventory; regulatory decisions and agreements; and analyses of criticality, packaging, storage, transportation, and system performance of a geological repository. 16 refs., 6 figs., 4 tabs.

  20. Development of metal cask for nuclear spent fuel

    International Nuclear Information System (INIS)

    Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S.

    2001-01-01

    It is one of the realistic solutions against increasing demand on interim storage of spent fuel assemblies arising from nuclear power plants in Japan to apply dual purpose (transport and storage) metal casks. Since 1980's Mitsubishi Heavy Industries, Ltd. (MHI) has been contributing to develop metal cask technologies for utilities, etc. in Japan, and have established transport and storage cask design ''MSF series'' which realizes higher payload and reliability for long term storage. MSF series transport and storage casks use various new design concepts and materials to improve thermal performance of the cask, structural integrity of the basket, durability of the neutron shielding material and so on. This paper summarizes an outline of the cask design that can accommodate BWR spent fuel assemblies as well as the new technologies applied to the design and fabrication. (author)

  1. Technical strategy for the management of INEEL spent nuclear fuel

    International Nuclear Information System (INIS)

    1997-03-01

    This report presents evaluations, findings, and recommendations of the Idaho National Engineering and Environmental Laboratory (INEEL) Spent Nuclear Fuel Task Team. The technical strategy developed by the Task Team includes stabilization, near term storage, packaging, transport, and ultimate disposal. Key issues identified and discussed include waste characterization, criticality, packaging, waste form performance, and special fuels. Current plans focus on onsite needs, and include three central elements: (1) resolution of near-term vulnerabilities, (2) consolidation of storage locations, and (3) achieving dry storage in transportable packages. In addition to the Task Team report, appendices contain information on the INEEL spent fuel inventory; regulatory decisions and agreements; and analyses of criticality, packaging, storage, transportation, and system performance of a geological repository. 16 refs., 6 figs., 4 tabs

  2. Immobilization in cement of ion exchange resins from Spanish nuclear reactors

    International Nuclear Information System (INIS)

    Huebra, A.G. de la; Murillo, R.; Ortiz, S.J.

    1990-01-01

    Ion exchange materials used at nuclear power plants can be immobilized in cements less expensive than polymer matrices. Cement solidification of spent ion exchange resins shows swelling and cracking troubles (during setting time, or of storage). The objective of this study was to select the types of cement that produce the best quality on immobilization of three kinds of resins and to set up cement formulations containing the maximum possible loading of resin. Four cements were selected to carried out the study. After a study of hydration-dehydration phenomena of ion exchange resins, a systematic work has been carried out on immobilization. Tests were performed to study compressive strength and underwater stability by changing water/cement ratio and resin/cement ratio. Mixtures made with water, cement and resin only were loaded with 10% by weight dry resin. Mixtures with higher loadings show poor workability. Tests were carried out by adding organic plasticizers and silica products to improve waste loading. Plasticizers reduced water demand and silica products permit the use of more water. Leaching tests have been performed at 40 O C. In conclusion Blast Furnace Slag is the best cement for immobilization of ion exchange resin both bead and powdered form for mechanical strength, stability and leaching

  3. Licensing of spent nuclear fuel dry storage in Russia

    International Nuclear Information System (INIS)

    Kislov, A.I.; Kolesnikov, A.S.

    1999-01-01

    The Federal nuclear and radiation safety authority of Russia (Gosatomnadzor) being the state regulation body, organizes and carries out the state regulation and supervision for safety at handling, transport and storage of spent nuclear fuel. In Russia, the use of dry storage in casks will be the primary spent nuclear fuel storage option for the next twenty years. The cask for spent nuclear fuel must be applied for licensing by Gosatomnadzor for both storage and transportation. There are a number of regulations for transportation and storage of spent nuclear fuel in Russia. Up to now, there are no special regulations for dry storage of spent nuclear fuel. Such regulations will be prepared up to the end of 1998. Principally, it will be required that only type B(U)F, packages can be used for interim storage of spent nuclear fuel. Recently, there are two dual-purpose cask designs under consideration in Russia. One of them is the CONSTOR steel concrete cask, developed in Russia (NPO CKTI) under the leadership of GNB, Germany. The other cask design is the TUK-104 cask of KBSM, Russia. Both cask types were designed for spent nuclear RBMK fuel. The CONSTOR steel concrete cask was designed to be in full compliance with both Russian and IAEA regulations for transport of packages for radioactive material. The evaluation of the design criteria by Russian experts for the CONSTOR steel concrete cask project was performed at a first stage of licensing (1995 - 1997). The CONSTOR cask design has been assessed (strength analysis, thermal physics, nuclear physics and others) by different Russian experts. To show finally the compliance of the CONSTOR steel concrete cask with Russian and IAEA regulations, six drop tests have been performed with a 1:2 scale model manufactured in Russia. A test report was prepared. The test results have shown that the CONSTOR cask integrity is guaranteed under both transport and storage accident conditions. The final stage of the certification procedure

  4. Development of Spent Fuel Examination Technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho Dong; Park, K. J.; Shin, H. S. (and others)

    2007-04-15

    For the official operation of ACPF Facility Attachment based on facility declared DIQ was issued by IAEA and officialized upon ROK government approval. This procedure gives an essential ground to negotiate Joint Determination between governments of ROK and US. For ACPF process material accountability a neutron coincidence counting system was developed and calibrated with Cf-252 source. Its performance test demonstrated that over-all counting efficiency was about 21% with random error, 1.5% against calibration source, which found to be satisfactory to the expected design specification. A calibration curve derived by MCNP code with relationship between ASNC doublet counts vs. neutron activity of Cm-244 showed calibration constant to be 2.78x10E5 counts/s.g which would be used for initial ACP hot operation test. Nuclear material transportation and temporary storage system was established for active demonstration of advanced spent fuel management process line and would be directly applied to the effective management of wastes arising from active demonstration and would later contribute as a base data to development of inter hot-cell movement system in pyro-processing line. In addition, an optimal spent fuel for the ACP demonstration was selected and a computer code was developed as a tool to estimate the expected source term at each key measurement point of ACP.

  5. Characteristics of spent nuclear fuel

    International Nuclear Information System (INIS)

    Notz, K.J.

    1988-04-01

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

  6. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix D, Part B: Naval spent nuclear fuel management

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This volume contains the following attachments: transportation of Naval spent nuclear fuel; description of Naval spent nuclear receipt and handling at the Expended Core Facility at the Idaho National Engineering Laboratory; comparison of storage in new water pools versus dry container storage; description of storage of Naval spent nuclear fuel at servicing locations; description of receipt, handling, and examination of Naval spent nuclear fuel at alternate DOE facilities; analysis of normal operations and accident conditions; and comparison of the Naval spent nuclear fuel storage environmental assessment and this environmental impact statement.

  7. 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)

  8. Spent nuclear fuel discharges from US reactors 1993

    Energy Technology Data Exchange (ETDEWEB)

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  9. Handling of final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    1978-01-01

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

  10. 49 CFR 173.173 - Paint, paint-related material, adhesives, ink and resins.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Paint, paint-related material, adhesives, ink and... Than Class 1 and Class 7 § 173.173 Paint, paint-related material, adhesives, ink and resins. (a) When..., paint-related material, adhesives, ink and resins must be packaged as follows: (1) As prescribed in...

  11. Resin secretory structures of Boswellia papyrifera and implications for frankincense yield

    NARCIS (Netherlands)

    Tolera, M.; Menger, D.; Sass, U.G.W.; Sterck, F.J.; Copini, P.; Bongers, F.

    2013-01-01

    Frankincense, a gum-resin, has been tapped from Boswellia papyrifera trees for centuries. Despite the intensive tapping and economic interest of B. papyrifera, information on the resin secretory structures, which are responsible for synthesis, storage and transport of frankincense, is virtually

  12. Thermal analysis for spent fuel casks

    International Nuclear Information System (INIS)

    Wells, A.H.

    1986-01-01

    Thermal analyses for spent fuel storage or transportation must demonstrate that the heat produced in the fuel will be removed without causing excessive fuel cladding temperatures or cask surface temperatures. The time required to ship fuel from a reactor to another site is a matter of days so temperature control during shipment is a short term concern and temperatures of 1100-1200 degrees F are acceptable in a hypothetical fire scenario, but storage is envisioned for twenty years or more and temperatures are limited to 380 degrees C (716 F). Elevated temperatures do more that weaken fuel cladding: the materials of the fuel basket also weaken with higher temperature, thermal stresses may cause cracks to form in welds, and impact limiter materials may soften or degrade. The high temperatures of a fire cause loss of liquid neutron shields and may cause solid materials to char and insulate a cask after the fire

  13. Potential Interface Issues in Spent Fuel Management

    International Nuclear Information System (INIS)

    2015-10-01

    This publication is an output of a series of meetings to identify and evaluate issues and opportunities associated with interfaces in the back end of the fuel cycle (BEFC) and to describe effective management approaches based on the experience of Member States. During the meetings, participants from Member States and other international organizations shared and evaluated the main interfaces and potential interface issues among the spent fuel storage, transport, reprocessing and disposal of the BEFC, and also reviewed the national approaches to addressing these issues. The aim of this publication is to provide an approach to identify the interfaces in the BEFC as well as the potential issues that should be addressed. It also aims at responding to the solutions Member States most often find to be effective and, in some cases, were adjusted or revisited to reach the fixed target. Most of the interfaces and issues are country specific, as evidenced by the variety and diversity of examples provided in this publication

  14. Multi-purpose container technologies for spent fuel management

    International Nuclear Information System (INIS)

    2000-12-01

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

  15. Adsorption of an Uranyl Carbonate Complex onto an Amidoxime Resin

    Energy Technology Data Exchange (ETDEWEB)

    Joe, Kih-Soo; Lee, Eil-Hee; Kim, Kwang-Wook; Jung, Euo-Chang; Jee, Kwang-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-10-15

    Uranium is a major element in the management of spent nuclear fuel or radioactive waste. Uranium is also present in sea water at a trace level ({approx}3ppb). Usually, an acid medium in a solvent extraction or ion exchange system have been used for the separation of uranium from a radioactive waste solution. Especially, some adsorbants containing amidoxime functional groups selective for the uranyl ion have been developed for the recovery of uranium in sea water. An amidoxime resin as an anion exchanger synthesized using divinylbenzene copolymer as a supporting material was applied for the separation of uranium and plutonium from simulated fission products in carbonate solution by making an uranyl tricarbonate complex, UO{sub 2}(CO{sub 3}){sub 3}{sup 4-}, in an alkaline solution of a carbonate (pH>8). A carbonate leaching of uranium from uranium contaminated soils was also successfully performed using a sodium carbonate and a sodium bicarbonate solution. In this work, the distribution coefficients of uranyl ions were measured for the amidoxime resins with a variation of carbonate concentration and the pH of the solutions as a preliminary work for a spent fuel treatment with a strong alkaline carbonate solution because this system has an advantage to reduce a high level waste volume by a selective oxidation and an adsorption of uranium from spent fuels.

  16. Flame Retardant Epoxy Resins

    Science.gov (United States)

    Thompson, C. M.; Smith, J. G., Jr.; Connell, J. W.; Hergenrother, P. M.; Lyon, R. E.

    2004-01-01

    As part of a program to develop fire resistant exterior composite structures for future subsonic commercial aircraft, flame retardant epoxy resins are under investigation. Epoxies and their curing agents (aromatic diamines) containing phosphorus were synthesized and used to prepare epoxy formulations. Phosphorus was incorporated within the backbone of the epoxy resin and not used as an additive. The resulting cured epoxies were characterized by thermogravimetric analysis, propane torch test, elemental analysis and microscale combustion calorimetry. Several formulations showed excellent flame retardation with phosphorous contents as low as 1.5% by weight. The fracture toughness of plaques of several cured formulations was determined on single-edge notched bend specimens. The chemistry and properties of these new epoxy formulations are discussed.

  17. Bismaleimide Copolymer Matrix Resins

    Science.gov (United States)

    Parker, John A.; Heimbuch, Alvin H.; Hsu, Ming-Ta S.; Chen, Timothy S.

    1987-01-01

    Graphite composites, prepared from 1:1 copolymer of two new bismaleimides based on N,N'-m-phenylene-bis(m-amino-benzamide) structure have mechanical properties superior to those prepared from other bismaleimide-type resins. New heat-resistant composites replace metal in some structural applications. Monomers used to form copolymers with superior mechanical properties prepared by reaction of MMAB with maleic or citraconic anhydride.

  18. System for removing contaminants from plastic resin

    Science.gov (United States)

    Bohnert, George W.; Hand, Thomas E.; DeLaurentiis, Gary M.

    2010-11-23

    A resin recycling system that produces essentially contaminant-free synthetic resin material in an environmentally safe and economical manner. The system includes receiving the resin in container form. A grinder grinds the containers into resin particles. The particles are exposed to a solvent in one or more solvent wash vessels, the solvent contacting the resin particles and substantially removing contaminants on the resin particles. A separator is used to separate the resin particles and the solvent. The resin particles are then placed in solvent removing element where they are exposed to a solvent removing agent which removes any residual solvent remaining on the resin particles after separation.

  19. Paramagnetic epoxy resin

    Directory of Open Access Journals (Sweden)

    E. C. Vazquez Barreiro

    2017-01-01

    Full Text Available This work illustrates that macrocycles can be used as crosslinking agents for curing epoxy resins, provided that they have appropriate organic functionalities. As macrocycles can complex metal ions in their structure, this curing reaction allows for the introduction of that metal ion into the resin network. As a result, some characteristic physical properties of the metallomacrocycle could be transferred to the new material. The bisphenol A diglycidyl ether (BADGE, n = 0 and hemin (a protoporphyrin IX containing the Fe(III ion, and an additional chloride ligand have been chosen. The new material has been characterized by differential scanning calorimetry (DSC, thermogravimetric analysis (TGA, Fourier Transform Infrared (FT-IR, Nuclear Magnetic Resonance (NMR, Transmission Electron Microscopy (TEM, and magnetic susceptibility measurements. Fe(III remains in the high-spin state during the curing process and, consequently, the final material exhibits the magnetic characteristics of hemin. The loss of the chlorine atom ligand during the cure of the resin allows that Fe(III can act as Lewis acid, catalyzing the crosslinking reactions. At high BADGE n = 0/hemin ratios, the formation of ether and ester bonds occurs simultaneously during the process.

  20. Metals removal from spent salts

    Science.gov (United States)

    Hsu, Peter C.; Von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Brummond, William A.; Adamson, Martyn G.

    2002-01-01

    A method and apparatus for removing metal contaminants from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents may be added to precipitate the metal oxide and/or the metal as either metal oxide, metal hydroxide, or as a salt. The precipitated materials are filtered, dried and packaged for disposal as waste or can be immobilized as ceramic pellets. More than about 90% of the metals and mineral residues (ashes) present are removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be spray-dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration ion exchange column, which yields salt solutions that contain less than 1.0 ppm of contaminants.

  1. Actinide removal from spent salts

    Science.gov (United States)

    Hsu, Peter C.; von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Adamson, Martyn G.

    2002-01-01

    A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration salt solutions that contain less than 0.1 ppm of thorium or uranium.

  2. TMI-2 spent fuel shipping

    International Nuclear Information System (INIS)

    Quinn, G.J.; Burton, H.M.

    1985-01-01

    TMI-2 failed fuel will be shipped to the Idaho National Engineering Laboratory for use in the DOE Core Examination Program. The fuel debris will be loaded into three types of canisters during defueling and dry loaded into a spent fuel shipping cask. The cask design accommodates seven canisters per cask and has two separate containment vessels with ''leaktight'' seals. Shipments are expectd to begin in early 1986

  3. Reuse of Hydrotreating Spent Catalyst

    International Nuclear Information System (INIS)

    Habib, A.M.; Menoufy, M.F.; Amhed, S.H.

    2004-01-01

    All hydro treating catalysts used in petroleum refining processes gradually lose activity through coking, poisoning by metal, sulfur or halides or lose surface area from sintering at high process temperatures. Waste hydrotreating catalyst, which have been used in re-refining of waste lube oil at Alexandria Petroleum Company (after 5 years lifetime) compared with the same fresh catalyst were used in the present work. Studies are conducted on partial extraction of the active metals of spent catalyst (Mo and Ni) using three leaching solvents,4% oxidized oxalic acid, 10% aqueous sodium hydroxide and 10% citric acid. The leaching experiments are conducting on the de coked extrude [un crushed] spent catalyst samples. These steps are carried out in order to rejuvenate the spent catalyst to be reused in other reactions. The results indicated that 4% oxidized oxalic acid leaching solution gave total metal removal 45.6 for de coked catalyst samples while NaOH gave 35% and citric acid gave 31.9 % The oxidized leaching agent was the most efficient leaching solvent to facilitate the metal removal, and the rejuvenated catalyst was characterized by the unchanged crystalline phase The rejuvenated catalyst was applied for hydrodesulfurization (HDS) of vacuum gas oil as a feedstock, under different hydrogen pressure 20-80 bar in order to compare its HDS activity

  4. Spent Fuel Working Group Report

    International Nuclear Information System (INIS)

    O'Toole, T.

    1993-11-01

    The Department of Energy is storing large amounts of spent nuclear fuel and other reactor irradiated nuclear materials (herein referred to as RINM). In the past, the Department reprocessed RINM to recover plutonium, tritium, and other isotopes. However, the Department has ceased or is phasing out reprocessing operations. As a consequence, Department facilities designed, constructed, and operated to store RINM for relatively short periods of time now store RINM, pending decisions on the disposition of these materials. The extended use of the facilities, combined with their known degradation and that of their stored materials, has led to uncertainties about safety. To ensure that extended storage is safe (i.e., that protection exists for workers, the public, and the environment), the conditions of these storage facilities had to be assessed. The compelling need for such an assessment led to the Secretary's initiative on spent fuel, which is the subject of this report. This report comprises three volumes: Volume I; Summary Results of the Spent Fuel Working Group Evaluation; Volume II, Working Group Assessment Team Reports and Protocol; Volume III; Operating Contractor Site Team Reports. This volume presents the overall results of the Working Group's Evaluation. The group assessed 66 facilities spread across 11 sites. It identified: (1) facilities that should be considered for priority attention. (2) programmatic issues to be considered in decision making about interim storage plans and (3) specific vulnerabilities for some of these facilities

  5. Biokompatibilitas Gelas Ionomer Modifikasi Resin

    OpenAIRE

    Rotua Lestari M

    2008-01-01

    Saat ini banyak berkembang material baru dalam dunia kedokteran gigi diantaranya adalah Gelas ionomer modifikasi resin yang dikembangkan untuk mengatasi kekurangan-kekurangan dari gelas ionomer konvensional. Adanya penambahan monomer resin daIam bentuk 2-hydroxyethylmetacylate (HEMA) telah meningkatkan kekuatan dari bahan ini. Gelas ionomer modifikasi resin mempunyai sifat-sifat fisis dan mekanis yang lebih baik dibandingkan dengan gelas ionomer konvensional. Gelas ionomer modifikasi ...

  6. Measurement of net nitrogen and phosphorus mineralization in wetland soils using a modification of the resin-core technique

    Science.gov (United States)

    Noe, Gregory B.

    2011-01-01

    A modification of the resin-core method was developed and tested for measuring in situ soil N and P net mineralization rates in wetland soils where temporal variation in bidirectional vertical water movement and saturation can complicate measurement. The modified design includes three mixed-bed ion-exchange resin bags located above and three resin bags located below soil incubating inside a core tube. The two inner resin bags adjacent to the soil capture NH4+, NO3-, and soluble reactive phosphorus (SRP) transported out of the soil during incubation; the two outer resin bags remove inorganic nutrients transported into the modified resin core; and the two middle resin bags serve as quality-control checks on the function of the inner and outer resin bags. Modified resin cores were incubated monthly for a year along the hydrogeomorphic gradient through a floodplain wetland. Only small amounts of NH4+, NO3-, and SRP were found in the two middle resin bags, indicating that the modified resin-core design was effective. Soil moisture and pH inside the modified resin cores typically tracked changes in the surrounding soil abiotic environment. In contrast, use of the closed polyethylene bag method provided substantially different net P and N mineralization rates than modified resin cores and did not track changes in soil moisture or pH. Net ammonification, nitrifi cation, N mineralization, and P mineralization rates measured using modified resin cores varied through space and time associated with hydrologic, geomorphic, and climatic gradients in the floodplain wetland. The modified resin-core technique successfully characterized spatiotemporal variation of net mineralization fluxes in situ and is a viable technique for assessing soil nutrient availability and developing ecosystem budgets.

  7. A mechanistic study of Protein A chromatography resin lifetime.

    Science.gov (United States)

    Jiang, Canping; Liu, Jing; Rubacha, Michael; Shukla, Abhinav A

    2009-07-31

    A mechanistic study into Protein A chromatographic resin lifetime limitations is presented. Binding and mass transport properties of two widely used agarose-based Protein A resins were studied to distinguish between the roles of resin fouling due to product/impurity build-up and ligand degradation as contributory factors towards the decline in binding capacity with use. Cycling studies were conducted with and without product loading on the columns to separate out the influence of resin fouling. Ligand degradation under the mildly alkaline conditions used for column regeneration was determined to be the primary cause for Protein A resin capacity decline with usage. The use of lower concentrations of caustic and the use of stabilizing excipients to protect the Protein A ligand during cleaning and sanitization were found to be useful techniques in maintaining column performance. The results presented in this paper provide a clearer understanding of the causative factors that limit Protein A chromatographic resin lifetime. It is anticipated that these findings will assist in the development of more robust and economical downstream manufacturing processes for monoclonal antibody and Fc fusion protein purification.

  8. Development of a computer program for the cost analysis of spent fuel management

    International Nuclear Information System (INIS)

    Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won; Cha, Jeong Hun; Whang, Joo Ho

    2009-01-01

    So far, a substantial amount of spent fuels have been generated from the PWR and CANDU reactors. They are being temporarily stored at the nuclear power plant sites. It is expected that the temporary storage facility will be full of spent fuels by around 2016. The government plans to solve the problem by constructing an interim storage facility soon. The radioactive management act was enacted in 2008 to manage the spent fuels safety in Korea. According to the act, the radioactive waste management fund which will be used for the transportation, interim storage, and the final disposal of spent fuels has been established. The cost for the management of spent fuels is surprisingly high and could include a lot of uncertainty. KAERI and Kyunghee University have developed cost estimation tools to evaluate the cost for a spent fuel management based on an engineering design and calculation. It is not easy to develop a tool for a cost estimation under the situation that the national policy on a spent fuel management has not yet been fixed at all. Thus, the current version of the computer program is based on the current conceptual design of each management system. The main purpose of this paper is to introduce the computer program developed for the cost analysis of a spent fuel management. In order to show the application of the program, a spent fuel management scenario is prepared, and the cost for the scenario is estimated

  9. Overview of spent fuel management and problems

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  10. Recent Progress on the Standardized DOE Spent Nuclear Fuel Canister

    International Nuclear Information System (INIS)

    Morton, D.K.; Snow, S.D.; Rahl, T.E.; Hill, T.J.; Morissette, R.P.

    2002-01-01

    The Department of Energy (DOE) has developed a set of containers for the handling, interim storage, transportation, and disposal in the national repository of DOE spent nuclear fuel (SNF). This container design, referred to as the standardized DOE SNF canister or standardized canister, was developed by the Department's National Spent Nuclear Fuel Storage Program (NSNFP) working in conjunction with the Office of Civilian Radioactive Waste Management (OCRWM) and the DOE spent fuel sites. This canister had to have a standardized design yet be capable of accepting virtually all of the DOE SNF, be placed in a variety of storage and transportation systems, and still be acceptable to the repository. Since specific design details regarding the storage, transportation, and repository disposal of DOE SNF were not finalized, the NSNFP recognized the necessity to specify a complete DOE SNF canister design. This allowed other evaluations of canister performance and design to proceed as well as providing standardized canister users adequate information to proceed with their work. This paper is an update of a paper presented to the 1999 American Nuclear Society of Mechanical Engineers (ASME) Pressure Vessels and Piping (PVP) Conference. It discusses recent progress achieved in various areas to enhance acceptance of this canister not only by the DOE complex but also fabricators and regulatory agencies

  11. A Study of Integrity Evaluation System for Spent Fuel and Selection of the Representative Spent Fuel

    International Nuclear Information System (INIS)

    Kim, J. G.; Lee, S. K.; Lim, C. J.; Kim, J. K.; Lee, S. J.

    2014-01-01

    Spent fuel (SF) integrity evaluation is a regulatory requirement that is described in 10 CFR 71(transportation) and 10 CFR 72(storage) of the U. S. NRC licensing requirement. NRC regulation states that retrievability of SF after storage should be ensured and SF integrity under the normal condition must be guaranteed during transportation and handling process that is entailed before/during/after the interim storage. And SF integrity evaluation under the hypothetical accident condition is a core technology element for an assessment of critical, shielding, and containment. In this paper, SF integrity evaluation system which is suitable for domestic situation is suggested, and necessity of representative SF selection and its method is described. The ultimate goal of the SF integrity evaluation is to evaluate a safety margin in case of transportation/ handling/storage of SFs. It means that retrievability of SF after storage should be assured and SF integrity must be guaranteed at normal condition in the process of transportation/handling accompanied before/during/after interim storage. In Korea, SF integrity evaluation system is not established up to date. Especially, representative SF selection technology that is essential to SF integrity evaluation has not been fulfilled. To overcome this situation effectively, the methodology and technology of an overseas agency need to be benchmarked. In this paper, an overseas SF integrity evaluation system is analyzed, and an evaluation system suitable for domestic situation is suggested. Also, necessity of representative SF selection and its method is described

  12. HTGR spent fuel storage study

    International Nuclear Information System (INIS)

    Burgoyne, R.M.; Holder, N.D.

    1979-04-01

    This report documents a study of alternate methods of storing high-temperature gas-cooled reactor (HTGR) spent fuel. General requirements and design considerations are defined for a storage facility integral to a fuel recycle plant. Requirements for stand-alone storage are briefly considered. Three alternate water-cooled storage conceptual designs (plug well, portable well, and monolith) are considered and compared to a previous air-cooled design. A concept using portable storage wells in racks appears to be the most favorable, subject to seismic analysis and economic evaluation verification

  13. [Acrylic resin removable partial dentures

    NARCIS (Netherlands)

    Baat, C. de; Witter, D.J.; Creugers, N.H.J.

    2011-01-01

    An acrylic resin removable partial denture is distinguished from other types of removable partial dentures by an all-acrylic resin base which is, in principle, solely supported by the edentulous regions of the tooth arch and in the maxilla also by the hard palate. When compared to the other types of

  14. Cure shrinkage in casting resins

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, J. Brock [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-02-01

    A method is described whereby the shrinkage of a casting resin can be determined. Values for the shrinkage of several resin systems in frequent use by Sandia have been measured. A discussion of possible methods for determining the stresses generated by cure shrinkage and thermal contraction is also included.

  15. Handling and transfer operations for partially-spent nuclear fuel

    International Nuclear Information System (INIS)

    Ibrahim, J.K.

    1983-01-01

    This project involved the handling and transfer of partially-spent reactor fuel from the Oregon State University TRIGA Reactor in Corvallis, Oregon to Hanford Engineering Development Laboratory in Richland, Washington. The method of handling is dependent upon the burn-up history of the fuel elements. Legal constraints imposed by standing U.S. nuclear regulations determine the selection of transport containers, transportation procedures, physical security arrangements in transit and nuclear material accountability documentation. Results of in-house safety evaluations of the project determine the extent of involvement of pertinent nuclear regulatory authorities. The actual handling activities and actual radiation dose rates are also presented (author)

  16. Intermodal transfer of spent fuel

    International Nuclear Information System (INIS)

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

    1993-01-01

    This paper discusses RADTRAN calculational models and parameter values for describing dose to workers during incident-free ship-to-truck transfer of spent fuel. Data obtained during observation of the offloading of research reactor spent fuel at Newport News Terminal in the Port of Hampton Roads, Virginia, are described. These data include estimates of exposure times and distances for handlers, inspectors, and other workers during offloading and overnight storage. Other workers include crane operators, scale operators, security personnel, and truck drivers. The data are compared to the default data in RADTRAN 4, and the latter are found to be conservative. The casks were loaded under IAEA supervision at their point of origin, and three separate radiological inspections of each cask were performed at the entry to the port (Hampton Roads) by the U.S. Coast Guard, the state of Virginia, and the shipping firm. As a result of the international standardization of containerized cargo handling in ports around the world, maritime shipment handling is particularly uniform. Thus, handler exposure parameters will be relatively constant for ship-truck and ship-rail transfers at ports throughout the world. Inspectors' doses are expected to vary because of jurisdictional considerations. The results of this study should be applicable to truck-to-rail transfers. (author)

  17. Criticality safety assessment of WWER-1000 spent fuel cask

    International Nuclear Information System (INIS)

    Apostolov, T.; Manolova, M.; Prodanova, R.

    2001-01-01

    A methodology that allows taking credit for burnup in the criticality safety analysis of WWER spent fuel casks is presented. It is based on the two world well known and used code systems:NESSEL-NUKO for depletion and SCALE-4.4 for criticality calculations. The results of criticality calculations of WWER-1000 spent fuel storage and transportation cask, applying burnup credit is shown. The depletion calculations have been carried out for three types of WWER-1000 fuel assemblies (with enrichment of 3.0%, 3.3% and 3.3% profiled) by modelling the real operational history of the first three fuel cycles at unit 6, Kozloduy NPP. The criticality calculational model has been developed on the basis of real fuel cask, designed by the Izorskie zavody. The results obtained show that the criticality safety criterion K eff < 0.95 is satisfied for both fresh and spent fuel. Besides, the implementation of burnup credit accounts for the reduced reactivity of spent fuel and allows evaluating the conservatism of the fresh fuel assumption. (author)

  18. Spent nuclear fuel storage device and spent nuclear fuel storage method using the device

    International Nuclear Information System (INIS)

    Tani, Yutaro

    1998-01-01

    Storage cells attachably/detachably support nuclear fuel containing vessels while keeping the vertical posture of them. A ventilation pipe which forms air channels for ventilating air to the outer circumference of the nuclear fuel containing vessel is disposed at the outer circumference of the nuclear fuel containing vessel contained in the storage cell. A shielding port for keeping the support openings gas tightly is moved, and a communication port thereof can be aligned with the upper portion of the support opening. The lower end of the transporting and containing vessel is placed on the shielding port, and an opening/closing shutter is opened. The gas tightness is kept by the shielding port, the nuclear fuel containing vessel filled with spent nuclear fuels is inserted to the support opening and supported. Then, the support opening is closed by a sealing lid. (I.N.)

  19. Spent fuel. Dissolution and oxidation

    International Nuclear Information System (INIS)

    Grambow, B.

    1989-03-01

    Data from studies of the low temperature air oxidation of spent fuel were retrieved in order to provide a basis for comparison between the mechanism of oxidation in air and corrosion in water. U 3 O 7 is formed by diffusion of oxygen into the UO 2 lattice. A diffusion coefficient of oxygen in the fuel matric was calculated for 25 degree C to be in the range of 10 -23 to 10 -25 m 2 /s. The initial rates of U release from spent fuel and from UO 2 appear to be similar. The lowest rates (at 25 degree c >10 -4 g/(m 2 d)) were observed under reducing conditions. Under oxidizing conditions the rates depend mainly of the nature and concentraion of the oxidant and/or on corbonate. In contact with air, typical initial rates at room temperature were in the range between 0.001 and 0.1 g/(m 2 d). A study of apparent U solubility under oxidizing conditions was performed and it was suggested that the controlling factor is the redox potential at the UO 2 surface rather than the E h of the bulk solution. Electrochemical arguments were used to predict that at saturation, the surface potential will eventually reach a value given by the boundaries at either the U 3 O 7 /U 3 O 8 or the U 3 O 7 /schoepite stability field, and a comparison with spent fuel leach data showed that the solution concentration of uranium is close to the calculated U solubility at the U 3 O 7 /U 3 O 8 boundary. The difference in the cumulative Sr and U release was calculated from data from Studsvik laboratory. The results reveal that the rate of Sr release decreases with the square root of time under U-saturated conditions. This time dependence may be rationalized either by grain boundary diffusion or by diffusion into the fuel matrix. Hence, there seems to be a possibility of an agreement between the Sr release data, structural information and data for oxygen diffusion in UO 2 . (G.B.)

  20. Experience with failed or damaged spent fuel and its impacts on handling

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1989-12-01

    Spent fuel management planning needs to include consideration of failed or damaged spent light-water reactor (LWR) fuel. Described in this paper, which was prepared under the Commercial Spent Fuel Management (CSFM) Program that is sponsored by the US Department of Energy (DOE), are the following: the importance of fuel integrity and the behavior of failed fuel, the quantity and burnup of failed or damaged fuel in storage, types of defects, difficulties in evaluating data on failed or damaged fuel, experience with wet storage, experience with dry storage, handling of failed or damaged fuel, transporting of fuel, experience with higher burnup fuel, and conclusions. 15 refs

  1. Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project (SPAR-II)

    International Nuclear Information System (INIS)

    2012-01-01

    As storage of spent fuel has become a key technology in spent fuel management, wet and dry storage have become mature technologies and continue to demonstrate good performance. Increased spent fuel storage capacity in combination with longer storage durations will be needed over the foreseeable future as many countries have delayed their decision on spent fuel disposal or reprocessing. Extended spent fuel storage is, and will remain, an important activity for all countries with nuclear power programmes. A number of countries are planning or have already initiated research programmes on spent fuel storage performance, and there is a continuing benefit in exchanging spent fuel storage experience of the Member States in order to build a comprehensive technology knowledge base. Potential degradation mechanisms that may affect cladding integrity during wet storage are uniform corrosion, pitting, galvanic, and microbiologically-influenced corrosion. Potential degradation mechanisms that may affect cladding integrity during dry storage and subsequent handling and transportation operations are air oxidation, thermal creep, stress corrosion cracking (SCC), delayed hydride cracking (DHC), hydride re-orientation, hydrogen migration and re-distribution. Investigations carried out so far indicate that from the degradation mechanisms that may affect the integrity of spent fuel assembly/bundle structure during interim storage, hydride re-orientation has the potential to impair the ability of the cladding to effectively withstand potentially adverse mechanical challenges resulting from handling or transportation accidents. Fuel integrity issues are related to the definition and criteria of fuel integrity, failure classification, packaging and retrieval of damaged fuel and transport of damaged fuel assemblies. Various monitoring technologies have been developed and used to confirm the continued spent fuel integrity during storage or to provide an early indication of developing

  2. Radioactive Source Specification of Bushehr’s VVER-1000 Spent Fuels

    OpenAIRE

    Rezaeian, Mahdi; Kamali, Jamshid

    2016-01-01

    Due to high radioactivity and significant content of medium- and long-lived radionuclides, different operations with spent nuclear fuels (e.g., handling, transportation, and storage) shall be accompanied by suitable radiation protections. On the other hand, determination of radioactive source specification is the initial step for any radiation protection design. In this study, radioactive source specification of the spent fuels of Bushehr nuclear power plant, which is a VVER-1000 type pressur...

  3. [Acrylic resin removable partial dentures].

    Science.gov (United States)

    de Baat, C; Witter, D J; Creugers, N H J

    2011-01-01

    An acrylic resin removable partial denture is distinguished from other types of removable partial dentures by an all-acrylic resin base which is, in principle, solely supported by the edentulous regions of the tooth arch and in the maxilla also by the hard palate. When compared to the other types of removable partial dentures, the acrylic resin removable partial denture has 3 favourable aspects: the economic aspect, its aesthetic quality and the ease with which it can be extended and adjusted. Disadvantages are an increased risk of caries developing, gingivitis, periodontal disease, denture stomatitis, alveolar bone reduction, tooth migration, triggering of the gag reflex and damage to the acrylic resin base. Present-day indications are ofa temporary or palliative nature or are motivated by economic factors. Special varieties of the acrylic resin removable partial denture are the spoon denture, the flexible denture fabricated of non-rigid acrylic resin, and the two-piece sectional denture. Furthermore, acrylic resin removable partial dentures can be supplied with clasps or reinforced by fibers or metal wires.

  4. Anion-exchange resin-based desulfurization process. Annual technical progress report, October 1, 1991--September 30, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Sheth, A.C.; Strevel, S.D.; Dharmapurikar, R.

    1992-12-31

    Under the current grant, the University of Tennessee Space Institute (UTSI) will carry out the bench scale evaluation and further development of the anion-exchange resin-based desulfurization concept to desulfurize alkali metal sulfates. This concept has been developed and patented by UTSI under US Patent No. 4,917,874. The developmental program proposed under this DOE grant includes screening of commercially available resins to select three candidate resins for further study. These three resins will undergo a series of experiments designed to test the resins` performance under different process conditions (including the use of spent MHD seed material). The best of these resins will be used in optimizing the regeneration step and in testing the effects of performance enhancers. The process schematic developed from the results will be used to estimate the related economics. During this reporting period, October 1, 1991 to September 30, 1992, analysis of batch mode screening experiments was completed to select three candidate resins for process variables study in the fixed-bed set-up. This setup was modified and the experiments were carded out to evaluate effects of major process variables. The analysis of fixed-bed experiments is going on and we have also started simple batch mode experiments to identify desirable conditions for resin regeneration step. We have also started simple process engineering type calculations to determine the trade-off between the solution concentration and the resulting evaporation/concentration load.

  5. Determination of degradation conditions of exchange resins containing technetium; Determinacion de condiciones de degradacion de resinas de intercambio conteniendo tecnecio

    Energy Technology Data Exchange (ETDEWEB)

    Rivera S, A.; Monroy G, F.; Quintero P, E., E-mail: aa_1190@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    The quantification of Tc-99 in spent exchange resins, coming from nuclear power plants, is indispensable to define their administration. The Tc-99 is a pure beta emitter of 210000 years of half-life, volatile and of a high mobility in water and soil. For this reason, the objective of this work is to establish a digestion method of ionic exchange resins containing technetium that retains more than 95% of this radioisotope. Mineralization tests were carried out of a resin Amberlite IRN-150 by means of an oxidation heat, in acid medium, varying the resin mass, the medium volume, the media type, the temperature and the digestion time. The digested samples were analyzed by gas chromatography to estimate the grade of their degradation. The {sup 99m}Tc was used as tracer to determine the technetium percentage recovered after mineralizing the resin. The digestion process depends on the temperature and the resin mass. At higher temperature better mineralization of samples and to greater resin mass to a constant temperature, less degradation of the resin. The spectra beta of the {sup 99m}Tc and {sup 99}Tc are presented. (Author)

  6. Licensing procedures for a dedicated ship for carrying spent nuclear fuel and radioactive waste. Report from workshop held at GOSAOMNADZOR, Moscow 2 -3 July 2001

    Energy Technology Data Exchange (ETDEWEB)

    Sneve, Margorzata K.; Bergman, Curt; Markarov, Valentin

    2001-07-01

    The report describes information exchange and discussion about the licensing principles and procedures for spent nuclear fuel and radioactive waste transportation at sea. Russian health, environment and safety requirements for transportation of waste by ships. (Author)

  7. Technical considerations associated with spent fuel acceptance. Final report

    International Nuclear Information System (INIS)

    Supko, E.M.

    1996-06-01

    This study was initiated by the Electric Power Research Institute (EPRI) to identify technical considerations associated with spent fuel acceptance and implementation of a waste management system that includes the use of transportable storage systems, and to serve as an opening dialogue among Standard Contract Holders and the department of Energy's Office of Civilian Radioactive Waste management (OCRWM) prior to the development of waste acceptance criteria or issuance of a Notice of Proposed Rulemaking by OCRWM to amend the Standard Contract. The original purpose of the Notice of Proposed Rulemaking was to address changes to the Standard Contract to implement a multi-purpose canister based system and to address other issues that were not adequately addressed in the standard contract. Even if DOE does not develop a multi-purpose canister based system for waste acceptance, it will still be necessary to develop waste acceptance criteria in order to accept spent fuel in transportable storage systems that are being deployed for at-reactor storage. In this study, technical issues associated with spent fuel acceptance will be defined and potential options and alternatives for resolution of technical considerations will be explored

  8. Chromatography resin support

    Science.gov (United States)

    Dobos, James G.

    2002-01-01

    An apparatus and method of using an improved chromatography resin support is disclosed. The chromatography support platform is provided by a stainless steel hollow cylinder adapted for being inserted into a chromatography column. An exterior wall of the stainless steel cylinder defines a groove for carrying therein an "O"-ring. The upper surface of the stainless steel column is covered by a fine stainless steel mesh welded to the edges of the stainless steel cylinder. When placed upon a receiving ledge defined within a chromatography column, the "O"-ring provides a fluid tight seal with the inner edge wall of the chromatography cylinder. The stainless steel mesh supports the chromatography matrix and provides a back flushable support which is economical and simple to construct.

  9. 21 CFR 872.3140 - Resin applicator.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Resin applicator. 872.3140 Section 872.3140 Food... DEVICES DENTAL DEVICES Prosthetic Devices § 872.3140 Resin applicator. (a) Identification. A resin applicator is a brushlike device intended for use in spreading dental resin on a tooth during application of...

  10. Spent fuel shipping cask accident evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Fields, S.R.

    1975-12-01

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

  11. Record of discussions in full wording: Hearing on the Ahaus storage facility for spent-fuel transport containers, June 21-29, 1983. Hearing concerning a project of the Deutsche Gesellschaft fuer Wiederaufarbeitung von Kernbrennstoffen mbH, Hannover, and STEAG Kernenergie GmbH, Essen, to establish a long-term storage facility for spent-fuel transport containers in Ahaus, Landkreis Borken, Nordrhein-Westfalen. Pt. 3

    International Nuclear Information System (INIS)

    1984-01-01

    This third part of the record of the Ahaus Hearing presents the full wording of the discussions and statements concerning the topics of radiation protection and protection of the environment during operation of the planned facility. The problems considered can be summarized under the following keynotes: Wastes, effluents, environmental monitoring, radiological protection of workers, micro-climate, accidents and their impacts on the environment, site selection, development trends, physical protection, emergency service. The final debates are concerned with the radwaste disposal and management concept of the Federal German Government, with alternative methods or techniques for waste storage, and with the transport of radwaste. The hearing was organized by the PTB in its capacity as a licensing authority under atomic energy law, and this organisation will be responsible of examining and evaluating the objections stated with a view to the requirements set by section 6 of the Atomic Energy Act. (HSCH) [de

  12. Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR-III) 2009–2014

    International Nuclear Information System (INIS)

    2015-10-01

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

  13. Regulation of spent nuclear fuel shipment: A state perspective

    International Nuclear Information System (INIS)

    Halstead, R.J.; Sinderbrand, C.; Woodbury, D.

    1987-01-01

    In 1985, the Wisconsin Department of Natural Resources (WDNR) sought to regulate rail shipments of spent nuclear fuel through the state, because federal regulations did not adequately protect the environmentally sensitive corridor along the route of the shipments. A state interagency working group identified five serious deficiencies in overall federal regulatory scheme: 1) failure to consider the safety or environmental risks associated with selected routes; 2) abscence of route-specific emergency response planning; 3) failure of the NRC to regulate the carrier of spent nuclear fuel or consider its safety record; 4) abscence of requirements for determination of need for, or the propriety of, specific shipments of spent nuclear fuel; and 5) the lack of any opportunity for meaningful public participation with respect to the decision to transport spent nuclear fuel. Pursuant to Wisconsin's hazardous substance statutes, the WDNR issues an order requiring the utility to file a spill prevention and mitigation plan or cease shipping through Wisconsin. A state trial court judge upheld the utility's challenge to Wisconsin's spill plan requirements, based on federal preemption of state authority. The state is now proposing federal legislation which would require: 1) NRC determination of need prior to approval of offsite shipment of spent fuel by the licensees; 2) NRC assessment of the potential environmental impacts of shipments along the proposed route, and comparative evaluation of alternative modes and routes; and 3) NRC approval of a route-specific emergency response and mitigation plan, including local training and periodic exercises. Additionally, the proposed legislation would authorize States and Indian Tribes to establish regulatory programs providing for permits, inspection, contingency plans for monitoring, containments, cleanup and decontamination, surveillance, enforcement and reasonable fees. 15 refs

  14. The first commercial spent fuel shipment in China

    International Nuclear Information System (INIS)

    Meinert, N.M.; Xiaoqing Li

    2004-01-01

    In two and a half years, government regulatory agencies and contractors from three countries worked together to design, license, fabricate, and transport the first commercial spent fuel shipment in China. Their cooperative efforts helped avoid the loss of full core reserve at a nuclear power plant serving two of China's largest cities. In March 2001, Everclean Environmental Engineering Corporation (EEEC) selected NAC International (NAC) to supply two United States Nuclear Regulatory Commission (USNRC) licensed Storable Transport Casks (NACSTC) and technology support, to ensure that qualified Chinese operators would be ready to load the first cask in late 2003. EEEC is a subsidiary of China National Nuclear Corporation (CNNC), which sets nuclear policy in China. EEEC is responsible for implementing nuclear transportation policy set forth from its parent corporation. Timely implementation of EEEC's ambitious plan would avoid loss of full core reserve at Guangdong Nuclear Power Station (Daya Bay) Unit-1, which supplies power to Hong Kong and Schenzen. The spent fuel would be transported to the Lanzhou Nuclear Fuel Complex (LNFC), a reprocessing facility, approximately 4,000 kilometers Northwest of Daya Bay

  15. Spent Nuclear Fuel Project operational staffing plan

    International Nuclear Information System (INIS)

    Debban, B.L.

    1996-03-01

    Using the Spent Nuclear Fuel (SNF) Project's current process flow concepts and knowledge from cognizant engineering and operational personnel, an initial assessment of the SNF Project radiological exposure and resource requirements was completed. A small project team completed a step by step analysis of fuel movement in the K Basins to the new interim storage location, the Canister Storage Building (CSB). This analysis looked at fuel retrieval, conditioning of the fuel, and transportation of the fuel. This plan describes the staffing structure for fuel processing, fuel movement, and the maintenance and operation (M ampersand O) staffing requirements of the facilities. This initial draft does not identify the support function resources required for M ampersand O, i.e., administrative and engineering (technical support). These will be included in future revisions to the plan. This plan looks at the resource requirements for the SNF subprojects, specifically, the operations of the facilities, balances resources where applicable, rotates crews where applicable, and attempts to use individuals in multi-task assignments. This plan does not apply to the construction phase of planned projects that affect staffing levels of K Basins

  16. Nuclide inventories of spent fuels from light water reactors

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Okamoto, Tsutomu

    2012-02-01

    Accurate information on nuclide inventories of spent fuels from Light Water Reactors (LWRs) is important for evaluations of criticality, decay heat, radioactivity, toxicity, and so on, in the safety assessments of storage, transportation, reprocessing and waste disposal of the spent fuels. So, a lot of lattice burn-up calculations were carried out for the possible fuel specifications and irradiation conditions in Japanese commercial LWRs by using the latest nuclear data library JENDL-4.0 and a sophisticated lattice burn-up calculation code MOSRA-SRAC. As a result, burn-up changes of nuclide inventories and their possible ranges were clarified for 21 heavy nuclides and 118 fission products, which are important from the viewpoint of impacts to nuclear characteristics and nuclear fuel cycle and environment. (author)

  17. Bulk-Fill Resin Composites

    DEFF Research Database (Denmark)

    Benetti, Ana Raquel; Havndrup-Pedersen, Cæcilie; Honoré, Daniel

    2015-01-01

    the restorative procedure. The aim of this study, therefore, was to compare the depth of cure, polymerization contraction, and gap formation in bulk-fill resin composites with those of a conventional resin composite. To achieve this, the depth of cure was assessed in accordance with the International Organization...... for Standardization 4049 standard, and the polymerization contraction was determined using the bonded-disc method. The gap formation was measured at the dentin margin of Class II cavities. Five bulk-fill resin composites were investigated: two high-viscosity (Tetric EvoCeram Bulk Fill, SonicFill) and three low......-viscosity (x-tra base, Venus Bulk Fill, SDR) materials. Compared with the conventional resin composite, the high-viscosity bulk-fill materials exhibited only a small increase (but significant for Tetric EvoCeram Bulk Fill) in depth of cure and polymerization contraction, whereas the low-viscosity bulk...

  18. and phenol–formaldehyde resin

    Indian Academy of Sciences (India)

    formaldehyde resin (PFR) modified with tetraethylorthosilicate are investigated in detail. The chemical synthesis of PFR, its modification with nanometer- sized SiO2 particles created by sol–gel method and subsequent coating, enables a preparation of ...

  19. Bending characteristics of resin concretes

    Directory of Open Access Journals (Sweden)

    Ribeiro Maria Cristina Santos

    2003-01-01

    Full Text Available In this research work the influence of composition and curing conditions in bending strength of polyester and epoxy concrete is analyzed. Various mixtures of resin and aggregates were considered in view of an optimal combination. The Taguchi methodology was applied in order to reduce the number of tests, and in order to evaluate the influence of various parameters in concrete properties. This methodology is very useful for the planning of experiments. Test results, analyzed by this methodology, shown that the most significant factors affecting bending strength properties of resin concretes are the type of resin, resin content and charge content. An optimal formulation leading to a maximum bending strength was achieved in terms of material parameters.

  20. Epoxy hydantoins as matrix resins

    Science.gov (United States)

    Weiss, J.

    1983-01-01

    Tensile strength and fracture toughness of castings of the hydantoin resins cured with methylenedianiline are significantly higher than MY 720 control castings. Water absorption of an ethyl, amyl hydantoin formulation is 2.1 percent at equilibrium and Tg's are about 160 C, approximately 15 deg below the final cure temperature. Two series of urethane and ester-extended hydantoin epoxy resins were synthesized to determine the effect of crosslink density and functional groups on properties. Castings cured with methylenedianiline or with hexahydrophthalic anhydride were made from these compounds and evaluated. The glass transition temperatures, tensile strengths and moduli, and fracture toughness values were all much lower than that of the simple hydantoin epoxy resins. Using a methylene bishydantoin epoxy with a more rigid structure gave brittle, low-energy fractures, while a more flexible, ethoxy-extended hydantoin epoxy resin gave a very low Tg.