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Sample records for fuel depository basic

  1. Dry vault for spent fuel depository, basic outsets, operating results and safety of the ''CASCAD'' plant

    International Nuclear Information System (INIS)

    Mercier, J.P.; Bardelle, P.

    1994-01-01

    Reprocessing and recycling of fissile materials is the preferred approach to spent fuel management in France. However, a number of spent fuel elements from prototype and experimental nuclear reactors cannot be reprocessed in the existing industrial facilities, either because such facilities are booked to full capacity, or due to technical factors such as non standard nature of fuel or limited series of fuel. The CEA therefore built a facility in which spent fuel can be stored for a few decades (50 years), until favourable conditions prevail for its disposal. The main features of this project consist of a dry depository, which presents a low cost of operation, against a wet on which is more expensive due to the circulation and the continuous controls of the water. Therefore, this is a fair solution because the experimental fuels will present a rather low residual heat power after decay in the nuclear reactor. At this stage, it becomes possible to cool the fuel elements by a fully passive air circulation. This process allows a good efficiency without mechanical equipment and works all the better as the amount of heat to exhaust is great, in the limits of the design. However, we will see that this concept may be extended to a depository of standard spent fuel elements. This facility, known as ''CASCAD'' (shortening for CASemate (=vault) CADarache) started up in 1990, and received its first canister of fuel on May 29th 1990. This paper reviews the basic design data of the facility, outlines the main techniques used for its construction, draws the safety concepts and presents the first results determined by a looking-back over 4 years of working. (author). 1 fig

  2. Dry vault for spent fuel depository, basic outsets, operating results and safety of the 'CASCAD' plant

    International Nuclear Information System (INIS)

    Bardelle, Ph.; Mercier, J.P.

    1995-01-01

    CEA built a facility in which spent fuel can be stored for a few decades (50 years), until favourable conditions prevail for its disposal. The main features of this project consist of a dry depository, which presents a low cost of operation, against a wet one which is more expensive due to the circulation and the continuous controls of the water. Fuel elements are cooled by a fully passive air circulation. This process allows a good efficiency without mechanical equipment and works all the better as the amount of heat to exhaust is great, in the limits of the design. This facility, known as 'CASCAD' (shortening for CASemate (=vault) CADarache) started up in 1990, and received its first canister of fuel on May 29th 1990. The basic design data of the facility, are reviewed the main techniques used for its construction are outlined the safety concepts are drawn and the first results determined by a looking-back over 4 years of working are presented. (K.A.). 1 tab

  3. Dry vault for spent fuel depository. Basic outsets, operating results and safety of the ''CASCAD'' plant

    International Nuclear Information System (INIS)

    Bardelle, P.

    1994-01-01

    Reprocessing and recycling of fissile materials is the preferred approach to spent fuel management in France. However, a number of spent fuel elements from prototype and experimental nuclear reactors cannot be reprocessed in the existing industrial facilities, either because such facilities are booked to full capacity, or due to technical factors such as non standard nature of fuel or limited series of fuel. The CEA therefore built a facility in which spent fuel can be stored for a few decades (50 years), until favourable conditions prevail for its disposal. The main features of this project consist in a dry depositary, which presents a low cost of working, against a wet one which is more expensive due to the circulation and the continuous controls of the water. Therefore, this is a fair solution because the experimental fuels will present a rather low residual heat power after decay in the nuclear reactor. At this stage, it becomes possible to cool the fuel elements by a fully passive air circulation. This process allows a good efficiency without mechanical equipment and works all the better as the amount of heat to exhaust is great, in the limits of the design. However, we will see that this concept may be extended to a depository of standard spent fuel elements. This facility, known as ''CASCAD'' (shortening for CASemate (=vault) CADarache) started up in 1990, and received its first canister of fuel on May 29 th 1990. This paper reviews the basic design data of the facility, outlines the main techniques used for its construction, draws the safety concepts and presents the first results determined by a looking-back over 4 years of working. (author)

  4. Dry vault for spent fuel depository. Basic outsets, operating results and safety of the ''CASCAD'' plant

    International Nuclear Information System (INIS)

    Bardelle, P.; Mercier, J.P.

    1994-01-01

    The building includes a depository concrete cell equipped with 315 depository wells. The cooling of these wells is guaranteed by an air natural circulation. A handling cell completes the installation. The depository layout allows to guarantee the safety-criticality even in case of 9 MSK strength earthquake. The biological protections supplied by concrete walls, the lead glass window and the shielding doors has been calculated in order to limit to 2 μ Sv/h, the dose rate at the work place. No staff contamination or irradiation was observed

  5. Interim report on nuclear waste depository thermal analysis

    International Nuclear Information System (INIS)

    Altenbach, T.J.

    1978-01-01

    A thermal analysis of a deep geologic depository for spent nuclear fuel is being conducted. The TRUMP finite difference heat transfer code is used to analyze a 3-dimensional model of the depository. The model uses a unit cell consisting of one spent fuel canister buried in salt beneath a ventilated room in the depository. A base case was studied along with several parametric variations. It is concluded that this method is appropriate for analyzing the thermal response of the system, and that the most important parameter in determining the maximum temperatures is the canister heat generation rate. The effects of room ventilation and different depository media are secondary

  6. Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository. Baseline concept criteria specifications and mechanical failure probabilities

    International Nuclear Information System (INIS)

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

    1979-05-01

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

  7. Dry vault for spent fuel depository. Basic outsets, operating results and safety of the CASCAD plant

    International Nuclear Information System (INIS)

    Bardelle, P.

    1997-01-01

    Reprocessing and recycling of fissile materials is the preferred approach to spent fuel management in France. However, a number of spent fuel elements from prototype and experimental nuclear reactors cannot be reprocessed in the existing industrial facilities, either because such facilities are booked to full capacity, or due to technical factors such as non standard nature of fuel or limited series of fuel. The CEA therefore built a facility in which spent fuel can be stored for a few decades (50 years), until favourable conditions prevail for its disposal. The main features of this project consist in a dry repository, which presents a low cost of working, against a wet one which is more expensive due to the circulation and the continuous control of the water. Therefore, this is a fair solution because the experimental fuels will present a rather low residual heat power after decay in the nuclear reactor. At this stage, it becomes possible to cool the fuel elements by a fully passive air circulation. This process allows a good efficiency without mechanical equipment and works all the better as the amount of heat to exhaust is great, in the limits of the design. However, we will see that this concept may be extended to a repository of standard spent fuel elements. This facility, known as 'CASCAD' (= CASemate CADarache) started up in 1990, and received its first canister of fuel on May 29, 1990. This paper reviews the basic design data of the facility, outlines the main techniques used for its construction, draws the safety concept and presents the first results determined by a looking-back over 4 years of working. (author)

  8. Fuel Cell Vehicle Basics | NREL

    Science.gov (United States)

    Fuel Cell Vehicle Basics Fuel Cell Vehicle Basics Researchers are developing fuel cells that can be silver four-door sedan being driven on a roadway and containing the words "hydrogen fuel cell electric" across the front and rear doors. This prototype hydrogen fuel cell electric vehicle was

  9. Comparison of thermally induced and naturally occurring water-borne leakages from hard rock depositories for radioactive waste

    International Nuclear Information System (INIS)

    Bourke, P.J.; Robinson, P.C.

    1981-01-01

    The relative importance of thermally induced and naturally occurring flows of water as causes of leakage from hard rock depositories for radioactive wastes is assessed. Separate analyses are presented for involatile, high level waste from reprocessing of fuel and for plutonium contaminated waste from fabrication of fuel. The effects of varying the quantities of wastes, pre-burial storage and the shapes and depths of depositories are considered. It is concluded that for representative values of these variables, thermal flow will remain the major cause of leakage for long times after the burial of both types of waste. (Auth.)

  10. 77 FR 48204 - Minority Depository Institution Advisory Committee

    Science.gov (United States)

    2012-08-13

    ...] Minority Depository Institution Advisory Committee AGENCY: Office of the Comptroller of the Currency... renewal of the Charter of the OCC Minority Depository Institution Advisory Committee (MDIAC) is necessary... and future development of minority depository institutions, in accordance with the goals established...

  11. 78 FR 5871 - Minority Depository Institutions Advisory Committee

    Science.gov (United States)

    2013-01-28

    ...] Minority Depository Institutions Advisory Committee AGENCY: Office of the Comptroller of the Currency... Comptroller of the Currency (OCC) announces a meeting of the Minority Depository Institutions Advisory... be open to the public. Agenda items include a discussion of the status of the minority depository...

  12. 76 FR 71438 - Minority Depository Institutions Advisory Committee

    Science.gov (United States)

    2011-11-17

    ...] Minority Depository Institutions Advisory Committee AGENCY: Department of the Treasury, Office of the... Currency (OCC) has determined to carry on the work of the Minority Depository Institutions Advisory... and in the public interest in order for the OCC to preserve the present number of minority depository...

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

    International Nuclear Information System (INIS)

    Bhaskaran, G.; McCleery, J.E.

    1979-10-01

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

  14. 78 FR 79076 - Minority Depository Institutions Advisory Committee; Meeting

    Science.gov (United States)

    2013-12-27

    ...] Minority Depository Institutions Advisory Committee; Meeting AGENCY: Office of the Comptroller of the... Office of the Comptroller of the Currency (OCC) announces a meeting of the Minority Depository..., DC 20219. Agenda items include a discussion of the status of the minority depository institution...

  15. 7 CFR 277.7 - Cash depositories.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 4 2010-01-01 2010-01-01 false Cash depositories. 277.7 Section 277.7 Agriculture... FOOD STAMP AND FOOD DISTRIBUTION PROGRAM PAYMENTS OF CERTAIN ADMINISTRATIVE COSTS OF STATE AGENCIES § 277.7 Cash depositories. (a) The term “cash depositories” refers to banks or other institutions which...

  16. 78 FR 59425 - Minority Depository Institutions Advisory Committee Meeting

    Science.gov (United States)

    2013-09-26

    ...] Minority Depository Institutions Advisory Committee Meeting AGENCY: Office of the Comptroller of the... Office of the Comptroller of the Currency (OCC) announces a meeting of the Minority Depository... discussion of the status of the minority depository institution industry and current topics of interest to...

  17. Depository Accounting of Securities in the Ukrainian Stock Market Regulation System

    Directory of Open Access Journals (Sweden)

    Veriha Hanna V.

    2016-02-01

    Full Text Available The aim of the article is to study the procedures, technologies, prudential regulation of depository accounting of securities in the system of the Ukrainian stock market infrastructure and identify directions of its improvement. The article analyzes the dynamics of the number of licenses issued by types of professional activity in the stock market. The necessity for further improvement of mechanisms of the updated system of depository accounting of securities in Ukraine has been proved. There have been developed the following recommendations: to improve the unified rules of accounting and regulation support of the system of risk management of depository activity; develop tools for prudential regulation of depository activity and strengthen the control over fulfillment of prudential standards by the Central Depository and depository institutions; create the necessary conditions for the practical implementation of legal norms concerning establishment of clearing institutions and increase in the level of competition between depositaries; expand the correspondent relations of the Central Depository in relation to the establishment of international depositary relations for the liberalization of the international movement of securities; use segregated accounts providing the possibility of storage of client funds separately from the funds of the transfer bank to protect the capital of the issuer and investor from risks of any force majeure situations; mediate the movement of funds at implementing dividend payments through participants of the accounting system: issuer-the Central Depository-depository institution-depositor.

  18. 76 FR 68064 - Reserve Requirements of Depository Institutions

    Science.gov (United States)

    2011-11-03

    ... Reserve Requirements of Depository Institutions AGENCY: Board of Governors of the Federal Reserve System. ACTION: Final rule. SUMMARY: The Board is amending Regulation D, Reserve Requirements of Depository Institutions, to reflect the annual indexing of the reserve requirement exemption amount and the low reserve...

  19. 75 FR 65563 - Reserve Requirements of Depository Institutions

    Science.gov (United States)

    2010-10-26

    ... Requirements of Depository Institutions AGENCY: Board of Governors of the Federal Reserve System. ACTION: Final rule. SUMMARY: The Board is amending Regulation D, Reserve Requirements of Depository Institutions, to reflect the annual indexing of the reserve requirement exemption amount and the low reserve tranche for...

  20. 78 FR 66249 - Reserve Requirements of Depository Institutions

    Science.gov (United States)

    2013-11-05

    ... Requirements of Depository Institutions AGENCY: Board of Governors of the Federal Reserve System. ACTION: Final rule. SUMMARY: The Board is amending Regulation D, Reserve Requirements of Depository Institutions, to reflect the annual indexing of the reserve requirement exemption amount and the low reserve tranche for...

  1. 77 FR 65773 - Reserve Requirements of Depository Institutions

    Science.gov (United States)

    2012-10-31

    ... Requirements of Depository Institutions AGENCY: Board of Governors of the Federal Reserve System. ACTION: Final rule. SUMMARY: The Board is amending Regulation D, Reserve Requirements of Depository Institutions, to reflect the annual indexing of the reserve requirement exemption amount and the low reserve tranche for...

  2. Spent nuclear fuel storage - Basic concept

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  3. 75 FR 51169 - OTS Minority Depository Institutions Advisory Committee

    Science.gov (United States)

    2010-08-18

    ... DEPARTMENT OF THE TREASURY Office of Thrift Supervision [Docket ID: OTS-2010-0026] OTS Minority... (OTS). ACTION: Notice. SUMMARY: The Charter for the OTS Minority Depository Institutions Advisory... of the Treasury to announce the renewal of the OTS Minority Depository Institutions Advisory...

  4. Installation of depository for radioactive material in rocks

    International Nuclear Information System (INIS)

    Bergman, S.G.A.; Sagefors, K.I.; Aakesson, B.Aa.

    1985-01-01

    The rock outside the depository has a hollow space which is filled by elastoplastic material possible to deform. The solid body of the depository has a central vertical shaft and concentric vertical outer shafts. Between the shafts there are vertically oriented layers with tunnels for storage of waste. The tunnels slope down from the central shaft. (G.B.)

  5. Integrating Decentralized Indoor Evacuation with Information Depositories in the Field

    Directory of Open Access Journals (Sweden)

    Haifeng Zhao

    2017-07-01

    Full Text Available The lonelier evacuees find themselves, the riskier become their wayfinding decisions. This research supports single evacuees in a dynamically changing environment with risk-aware guidance. It deploys the concept of decentralized evacuation, where evacuees are guided by smartphones acquiring environmental knowledge and risk information via exploration and knowledge sharing by peer-to-peer communication. Peer-to-peer communication, however, relies on the chance that people come into communication range with each other. This chance can be low. To bridge between people being not at the same time at the same places, this paper suggests information depositories at strategic locations to improve information sharing. Information depositories collect the knowledge acquired by the smartphones of evacuees passing by, maintain this information, and convey it to other passing-by evacuees. Multi-agent simulation implementing these depositories in an indoor environment shows that integrating depositories improves evacuation performance: It enhances the risk awareness and consequently increases the chance that people survive and reduces their evacuation time. For evacuating dynamic events, deploying depositories at staircases has been shown more effective than deploying them in corridors.

  6. 78 FR 4349 - Records of Failed Insured Depository Institutions

    Science.gov (United States)

    2013-01-22

    ... with the definition of ``records'' in section 210(a)(16)(D) of the Dodd-Frank Wall Street Reform and...-day operations prior to its failure. \\4\\ 12 U.S.C. 5390(a)(16)(D), which defines ``records'' to mean... depository institution, but insured depository institutions often retain copies of reports of examination and...

  7. 77 FR 66361 - Reserve Requirements of Depository Institutions: Reserves Simplification

    Science.gov (United States)

    2012-11-05

    ... Requirements of Depository Institutions: Reserves Simplification AGENCY: Board of Governors of the Federal... (Reserve Requirements of Depository Institutions) published in the Federal Register on April 12, 2012. The... simplifications related to the administration of reserve requirements: 1. Create a common two-week maintenance...

  8. 77 FR 21846 - Reserve Requirements of Depository Institutions: Reserves Simplification

    Science.gov (United States)

    2012-04-12

    ... Requirements of Depository Institutions: Reserves Simplification AGENCY: Board of Governors of the Federal Reserve System. ACTION: Final rule. SUMMARY: The Board is amending Regulation D, Reserve Requirements of Depository Institutions, to simplify the administration of reserve requirements. The final rule creates a...

  9. 75 FR 29386 - Consumer Protections for Depository Institution Sales of Insurance

    Science.gov (United States)

    2010-05-25

    ... DEPARTMENT OF THE TREASURY Office of Thrift Supervision Consumer Protections for Depository...: Consumer Protection for Depository Institution Sales of Insurance. OMB Number: 1550-0106. Regulation..., Federal Deposit Insurance Corporation, and Office of Thrift Supervision to prescribe joint consumer...

  10. 78 FR 56706 - Depository Library Council to the Public Printer; Meeting

    Science.gov (United States)

    2013-09-13

    ... GOVERNMENT PRINTING OFFICE Depository Library Council to the Public Printer; Meeting The Depository Library Council to the Public Printer (DLC) will meet on Monday, October 21, 2013 through.... Davita Vance-Cooks, Public Printer of the United States. [FR Doc. 2013-22247 Filed 9-12-13; 8:45 am...

  11. 75 FR 61760 - Depository Library Council to the Public Printer; Meeting

    Science.gov (United States)

    2010-10-06

    ... GOVERNMENT PRINTING OFFICE Depository Library Council to the Public Printer; Meeting The Depository Library Council to the Public Printer (DLC) will meet on Monday, October 18, 2010, through... regulations. Robert C. Tapella, Public Printer of the United States. [FR Doc. 2010-25047 Filed 10-5-10; 8:45...

  12. 77 FR 58381 - Depository Library Council to the Public Printer; Meeting

    Science.gov (United States)

    2012-09-20

    ... GOVERNMENT PRINTING OFFICE Depository Library Council to the Public Printer; Meeting The Depository Library Council to the Public Printer will meet on Monday, October 15, 2012 through Thursday... Printer of the United States. [FR Doc. 2012-23015 Filed 9-19-12; 8:45 am] BILLING CODE 1520-01-P ...

  13. 76 FR 13617 - Depository Library Council to the Public Printer Meeting

    Science.gov (United States)

    2011-03-14

    ... GOVERNMENT PRINTING OFFICE Depository Library Council to the Public Printer Meeting The Depository Library Council to the Public Printer (DLC) will meet on Monday, April 4, 2011 through Wednesday, April 6... Disabilities Act and meets all Fire Safety Act regulations. William J. Boarman, Public Printer of the United...

  14. 75 FR 18209 - Depository Library Council to the Public Printer; Meeting

    Science.gov (United States)

    2010-04-09

    ... GOVERNMENT PRINTING OFFICE Depository Library Council to the Public Printer; Meeting The Depository Library Council to the Public Printer (DLC) will meet on Monday, April 26, 2010, through Wednesday.... Robert C. Tapella, Public Printer of the United States. [FR Doc. 2010-8123 Filed 4-8-10; 8:45 am] BILLING...

  15. 76 FR 58005 - Meeting Notice; Depository Library Council to the Public Printer

    Science.gov (United States)

    2011-09-19

    ... GOVERNMENT PRINTING OFFICE Meeting Notice; Depository Library Council to the Public Printer The Depository Library Council to the Public Printer (DLC) will meet on Monday, October 17, through Thursday..., Public Printer of the United States. [FR Doc. 2011-23948 Filed 9-16-11; 8:45 am] BILLING CODE 1520-01-P ...

  16. The far field heating effects of a radioactive waste depository in hard rock

    International Nuclear Information System (INIS)

    Hodgkinson, D.P.; Bourke, P.J.

    1978-01-01

    Fission product heating of the rock surrounding a depository for high level radioactive waste, will result in high temperatures and high thermal gradients over distances of several hundred metres for many centuries. The consequent thermal expansion of the rock leads to stresses which could alter the fracture pattern and therefore the permeability of the rock. These problems are assessed by considering an idealised model of a depository for which analytic solutions to the temperature and stress fields are derived. A related problem is that any water present in the fissures will tend to rise because of its decrease in density on heating. If the water had previously leached away some of the radionuclides in the waste, then this convective transport constitutes a possible leakage path back to the biosphere. For the low permeabilities expected at a depository site, it is possible to linearise the resulting equations and derive analytic solutions for the flow velocities. This procedure has been carried out for the idealised depository model, in order to estimate the magnitude of these effects

  17. A basic research on the transient behavior for a metallic fuel FBR

    International Nuclear Information System (INIS)

    Baba, Mamoru; Hirano, Go; Kawada, Ken-ichi; Niwa, Hajime

    1999-03-01

    A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'A basic research on the transient behavior for a metallic fuel FBR'. The results are the following. (1) Target and Results of analysis: The accident initiator considered is a LOF accident without scram. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transition phase. It is shown that the moderator is quite elective to mitigate the accident at the initiation phase. However, it is necessary to analyze the transition phase to know if the re-criticality is totally avoided after the initiation phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large scale material movement in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics, and examined

  18. The content and environmental impact from the waste depository in Sillamaee

    International Nuclear Information System (INIS)

    Ehdwall, H.; Sundblad, B.; Nosov, V.; Putnik, H.; Mustonen, R.; Salonen, L.; Qvale, H.

    1994-01-01

    The studies of the waste depository in Sillamaee, Estonia, shows that the content as well as the wall material is typical tailings from chemical enrichment of uranium ore. The environmental impact from radioactive substances as well as heavy metals has been estimated. Results show the major radiological impact to the population in the Sillamaee town is the exposure to radon and its daughter products emanating from the depository. The impact on the Gulf of Finland is limited to a few hundred metres outside the coast-line. 7 refs, 8 figs, 15 tabs

  19. The basic research on the CDA initiation phase for a metallic fuel FBR

    International Nuclear Information System (INIS)

    Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime

    1998-03-01

    A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku University and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled. (1) Target and Results of analysis: The accident initiator considered is a LOF accident with ATWS. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transient phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large material movement to in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics. The results of a sample case, that is a metallic fueled core at the beginning of cycle, show this improvement is appropriate. (3) Conclusion: The behavior at CDA of a metallic fueled core of a fast reactor was analyzed using the CDA initiation phase analysis code and the knowledge of the important characteristics at the CDA initiation phase was obtained

  20. Radioactivity in soil and air near the Sillamaee tailings depository

    International Nuclear Information System (INIS)

    Realo, E.

    2002-01-01

    Radiological situation outside the Sillamaee plant and U/Th tailings depository is studied by soil monitoring and modelling methods. In the Sillamaee region, the reliable monitoring and assessment of the technological impact to public exposure is significantly interfered with the complex nature of natural background (e.g., elevated and variable soil concentrations of 226 Ra in the range of 40 .. 320 Bq kg -1 , high outdoor 222 Rn levels, etc.). The releases of radionuclides from the tailing depository to the ground and tap water are negligible. The major water pathway is the release to the Gulf of Finland and specifically the consumption of fish. The isotopes 234 U, 238 U (about 6.2 GBq a -1 or 0.025 % of the total U per year) and 210 Pb (∼ 0.05 GBq a -1 ) dominate in the source term. A compartmental modelling of this pathway has demonstrated a low radiological impact to the population with the individual and collective committed doses of 1 μSv and 1 manSv, respectively, for both current leakage and sudden dam collapse scenarios. We have determined the depth distribution of radionuclides in soil profiles to find an evidence for airborne radionuclide deposition. Samples collected in the Sillamaee town (up to 1 .. 2 km from the depository) have been analysed by using low background Ge(Li) and HPGe gamma spectrometry. The results demonstrate an enhanced (relative to the bottom layers) 226 Ra and 238 U content in the surface soil layers. A similar observation has been found near the oil-shale-fired power plants in NE Estonia, which we attributed to the long term deposition of fly-ash radionuclides. It should be noted that in the non-polluted locations over Estonia the 226 Ra concentration in the uppermost 0 .. 3 cm layer is about 11 % less than in the next 3 cm thick soil layers. In the ground level air, the mean dust loads of 0.23 mg m -3 (0.02 to 1.6 mg m -3 ) have been measured (V. Nossov, SILMET, 1997, private communication and [6]) with the aerosol beta

  1. Present state of the design and realization of regional radioactive waste depositories and waste acceptance criteria for disposal

    International Nuclear Information System (INIS)

    Kortus, J.

    1988-01-01

    Surface type regional depositories for radioactive wastes from nuclear power plants are described in detail. The depository of the Mochovce nuclear power plant is located near the plant, that of the Dukovany nuclear power plant is directly on the premises of the plant. Particular attention is paid to the design of the monolithic reinforced concrete pits, draining of rainwater from their surface, draining of seeping rainwater from the pit environment by means of a double drainage system, and insulation of the pits against water. The construction of the Mochovce depository started in 1987; some experience gained from this activity is presented. The radioactive waste acceptance criteria for depositories of this kind, based on safety analysis, are given. (author). 2 figs

  2. 31 CFR 560.319 - United States depository institution.

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false United States depository institution. 560.319 Section 560.319 Money and Finance: Treasury Regulations Relating to Money and Finance... entity, that is engaged primarily in the business of banking (for example, banks, savings banks, savings...

  3. 31 CFR 537.319 - U.S. depository institution.

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false U.S. depository institution. 537.319 Section 537.319 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE..., that is engaged primarily in the business of banking (for example, banks, savings banks, savings...

  4. 31 CFR 538.317 - U.S. depository institution.

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false U.S. depository institution. 538.317 Section 538.317 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE..., that is engaged primarily in the business of banking (for example, banks, savings banks, savings...

  5. Librarian-Initiated Publications Discovery: How Do Digital Depository Librarians Discover and Select Web-Based Government Publications for State Digital Depositories?

    Science.gov (United States)

    Lin, Chi-Shiou; Eschenfelder, Kristin R.

    2010-01-01

    This paper reports on a study of librarian initiated publications discovery (LIPD) in U.S. state digital depository programs using the OCLC Digital Archive to preserve web-based government publications for permanent public access. This paper describes a model of LIPD processes based on empirical investigations of four OCLC DA-based digital…

  6. Fuel cycle services

    International Nuclear Information System (INIS)

    Gruber, Gerhard J.

    1990-01-01

    TRIGA reactor operators are increasingly concerned about the back end of their Fuel Cycle due to a new environmental policy in the USA. The question how to close the Fuel Cycle will have to be answered by all operators sooner or later. Reprocessing of the TRIGA fuel elements is not available. Only long term storage and final disposal can be considered. But for such a storage or disposal a special treatment of the fuel elements and of course a final depository is necessary. NUKEM plans to undertake efforts to assist the TRIGA operators in this area. For that reason we need to know your special needs for today and tomorrow - so that potential processors can consider whether to offer these services on the market. (orig.)

  7. 75 FR 51168 - Consumer Protections for Depository Institution Sales of Insurance

    Science.gov (United States)

    2010-08-18

    ... addresses: Office of Information and Regulatory Affairs, Attention: Desk Officer for OTS, U.S. Office of..., advertising, and offers of any insurance product by a depository institution or by other persons performing...

  8. 75 FR 45204 - Consumer Protections for Depository Institution Sales of Insurance

    Science.gov (United States)

    2010-08-02

    ... addresses: Office of Information and Regulatory Affairs, Attention: Desk Officer for OTS, U.S. Office of..., solicitations, advertising, and offers of any insurance product by a depository institution or by other persons...

  9. Report of 5th new nuclear fuel research meeting, Yayoi Research Group. Trend of advanced basic research in nuclear fuel technical development

    International Nuclear Information System (INIS)

    1994-03-01

    Theme of this meeting is 'Trend of advanced basic research in nuclear fuel technical development', and it was attempted to balance both sides of the basic research and the development. At the meeting, lectures were given on the chemical form of FPs in oxide fuel pins, the absorption of hydrogen of fuel cladding tubes, the application of hydride fuel to thorium cycle, the thermal properties of fuel cladding tubes, the preparation of NpN and heat conductivity, the high temperature chemical reprocessing of nitride fuel, the research on the annihilation treatment of minor actinide in fast reactors, the separation of TRU by dry process and the annihilation using a metallic fuel FBR. In this report, the summaries of the lectures are collected, and also the program of the meeting and the list of attendants are shown. (K.I.)

  10. Leakage and accidental releases from Sillamaee waste depository and environmental risk assessment

    International Nuclear Information System (INIS)

    Bergstroem, U.; Nordlinder, S.; Aggeryd, I.

    1994-01-01

    An environmental risk assessment has been performed for the Sillamaee depository. The object of the study is to illustrate the consequences to man if the hazardous substances contained in the depository reach the sea. Two cases were studied: 1) the environmental impact at present conditions with a relatively low continuous leakage of the elements from the dam, and 2) the impact from a sudden dam collapse. The radiological dose impact on the population in the Baltic Sea area has been calculated for the observed leaking rate. The highest individual dose is less than 1μSv while for the dam collapse the dose will be in the order of 2 μSv and the dominant exposure pathway is via consumption of fish. The collective dose is about 1 manSv (emanating from fish caught in all parts of the Baltic Sea for both cases). Furthermore, the consequences of release of some metals (copper, zinc, niobium and molybdenum) were studied in the case of a dam break. In the Bay of Narva, outside the nearest coast, the additional contribution to the natural concentration in water will be neglectable for zinc and niobium. However, for copper and molybdenum the concentration will rise considerably during the first year. The additional load from the depository will still after 50 years be in the same order as published concentrations in the sea. The intake by man of those metals via fish caught in the Bay of Narva will be well below the limits of intake for zinc and molybdenum, (no recommendation of limits for niobium was found), while the calculated intake of copper from the depository will be in the same order as internationally recommended limits of intake. 34 refs

  11. Relationship between basic nuclear data and LWR fuel cycle parameters

    International Nuclear Information System (INIS)

    Becker, M.; Harris, D.R.; Quan, B.; Ryskamp, J.M.

    1979-01-01

    An interactive system has been developed at RPI to analyze the sensitivity of water reactor fuel cycle parameters and costs to uncertainties in nuclear data. A sequence of batch depletion, core analysis, and fuel cost codes (referred to as Path B) determines the changes in fuel cycle parameters and costs for changes in few-group microscopic cross sections, in fission yields, and in decay data. For cases that are found to be significant from Part B analysis, the sensitivities of few-group data to basic nuclear data are determined by detailed calculations (referred to as Path A). Analyses of pressurized and boiling water reactors with recycle and throwaway options show substantial sensitivities of fuel cycle parameters and costs, particularly to thermal and resonance nuclear data for fissile nuclides. The results bring out the importance for power reactor sensitivity analysis of dealing with the full fuel cycle including depletion of initially-loaded fuel and the building-in of actinides and fission products

  12. The M2 slowdown and depository intermediation: implications for monetary policy

    OpenAIRE

    John B. Carlson; Katherine A. Samolyk

    1992-01-01

    An examination of credit flow rechanneling away from depository institutions over the past decade in response to evolving financial markets and regulatory structure, and a discussion of how this trend has complicated monetary policymaking.

  13. Radioactive waste and the back part of fuel cycle of nuclear installations in Slovakia

    International Nuclear Information System (INIS)

    Koprda, V.

    2004-01-01

    This article is devoted to radioactive waste (RAW) management, an integrated system starting with collection and sorting of RAW through its storage, treatment, conditioning, handling and transport up to its disposal. Some notes will touch also the near surface depository of low level and intermediate level radioactive waste in Mochovce, and the long-term storage of waste improper for such type of disposal, and also some words will be addressed to the development and research of a deep geological depository for disposal spent fuel from nuclear power plant and long-lived radioactive waste. (author)

  14. Basic research and industrialization of CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su [and others

    2000-04-01

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  15. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su

    2000-04-01

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  16. 76 FR 709 - Electronic Funds Transfer of Depository Taxes; Correction

    Science.gov (United States)

    2011-01-06

    ... DEPARTMENT OF THE TREASURY Internal Revenue Service 26 CFR Parts 40 and 301 [TD 9507] RIN 1545-BJ13 Electronic Funds Transfer of Depository Taxes; Correction AGENCY: Internal Revenue Service (IRS...) providing guidance relating to Federal tax deposits (FTDs) by Electronic Funds Transfer (EFT). The temporary...

  17. 76 FR 708 - Electronic Funds Transfer of Depository Taxes; Correction

    Science.gov (United States)

    2011-01-06

    ... DEPARTMENT OF THE TREASURY Internal Revenue Service 26 CFR Parts 1, 31, 40, and 301 [TD 9507] RIN 1545-BJ13 Electronic Funds Transfer of Depository Taxes; Correction AGENCY: Internal Revenue Service... Electronic Funds Transfer (EFT). The temporary and final regulations provide rules under which depositors...

  18. Nonlibrary Partnerships: Acceptable Use (and Behavior) in the Web-based Depository.

    Science.gov (United States)

    Brinkerhoff, Kathie

    2000-01-01

    Discusses problems facing federal depository libraries regarding patron behavior when using Web sites. Topics include acceptable use policies; existing laws and enforcement offices; protecting library property; pornography, including child pornography; sexual harassment; and other forms of staff harassment. (LRW)

  19. Basic Research Needs for Clean and Efficient Combustion of 21st Century Transportation Fuels

    Energy Technology Data Exchange (ETDEWEB)

    McIlroy, A.; McRae, G.; Sick, V.; Siebers, D. L.; Westbrook, C. K.; Smith, P. J.; Taatjes, C.; Trouve, A.; Wagner, A. F.; Rohlfing, E.; Manley, D.; Tully, F.; Hilderbrandt, R.; Green, W.; Marceau, D.; O' Neal, J.; Lyday, M.; Cebulski, F.; Garcia, T. R.; Strong, D.

    2006-11-01

    To identify basic research needs and opportunities underlying utilization of evolving transportation fuels, with a focus on new or emerging science challenges that have the potential for significant long-term impact on fuel efficiency and emissions.

  20. BASIC Programming for the Integration of Money, Demand Deposits Creation, and the Hicksian-Keynesian Model.

    Science.gov (United States)

    Tom, C. F. Joseph

    Money, banking, and macroeconomic textbooks traditionally present the topics of money, the creation of demand deposits by depository institutions, and the Hicksian-Keynesian Theory of Income and Interest separately, as if they were unrelated. This paper presents an integrated approach to those subjects using computer programs written in BASIC, the…

  1. Basic tendencies of restructured UO2 nuclear fuels fabrication industry for water-moderated reactors

    International Nuclear Information System (INIS)

    Makhova, V.A.; Bokshitskij, V.I.; Blinova, I.V.

    2002-01-01

    Processes of reformation and consolidation of firms and frontier nuclear fuels fabrication industry associated with processes of globalization and deregulation of electric power market are analyzed. Current state of nuclear fuel market and basic factors influenced on the market are presented. The role of nuclear fuel in increasing competition of NPP and fundamental directions of innovation action on the creation of perspective kinds of fuel were considered [ru

  2. Electronic Government Information and the Depository Library Program: Paradise Found?

    Science.gov (United States)

    Dugan, Robert E.; Cheverie, Joan F.

    1992-01-01

    The trend in the 1980s to provide an increasingly larger amounts of government information in electronic formats has led to concerns regarding the inclusion of electronic items in the depository library program (DLP). Inclusion of electronic formats in the DLP provides increased public access but also poses technical problems for the depository…

  3. Basic research in support of innovative fuels design for the Generation IV systems (F-BRIDGE project)

    International Nuclear Information System (INIS)

    Valot, Carole; Bertolus, Marjorie; Konings, Rudy; Somers, Joe; Groot, Sander de

    2010-01-01

    F-BRIDGE (Basic Research in support of Innovative Fuels Design for the GEN IV systems) is a 4-year project which started in 2008. It seeks to bridge the gap between basic research and technological applications for generation IV nuclear reactor systems. One of the challenges for the next generation of reactors is to significantly increase the efficiency in designing innovative fuels. The object of the F-BRIDGE project is to complement the empirical approach by a physically-based description of fuel and cladding materials to enable a rationalization of the design process and a better selection of promising fuel systems. Advanced modelling and separate effects experiments are carried out in order to obtain more exact physical descriptions of ceramic fuels and cladding, at relevant scales from the atomic to the macroscopic scale. Research is also focused on assessing and improving 'sphere-pac' fuel, a composite-ceramics concept which has shown promise. The project activities can be broken down into four main areas: (i) Basic research investigations using a multi-scale approach in both experimentation and modelling to enable the generation of missing basic data, the identification of relevant mechanisms and the development of appropriate models; (ii) Transfer between technological issues and basic research by bringing together within the same project materials scientists, engineers and end-users; (iii) Assessment of the drawbacks and benefits of the sphere-pac fuel application to various Generation IV systems; (iv) Education and training to promote research in the field of fuel materials, to ensure the exchange of results and ideas among the participants and to link the project with other related European or international initiatives. The project relies on the complementary expertise of 19 partners: nuclear and non nuclear research organisations, universities, a nuclear engineering company, as well as technology and project management consultancy small and medium

  4. Radiological survey of the covered and uncovered drilling mud depository.

    Science.gov (United States)

    Jónás, Jácint; Somlai, János; Csordás, Anita; Tóth-Bodrogi, Edit; Kovács, Tibor

    2018-08-01

    In petroleum engineering, the produced drilling mud sometimes contains elevated amounts of natural radioactivity. In this study, a remediated Hungarian drilling mud depository was investigated from a radiological perspective. The depository was monitored before and after a clay layer was applied as covering. In this study, the ambient dose equivalent rate H*(10) of the depository has been measured by a Scintillator Probe (6150AD-b Dose Rate Meter). Outdoor radon concentration, radon concentration in soil gas, and in situ field radon exhalation measurements were carried out using a pulse-type ionization chamber (AlphaGUARD radon monitor). Soil gas permeability (k) measurements were carried out using the permeameter (RADON-JOK) in situ device. Geogenic radon potentials were calculated. The radionuclide content of the drilling mud and cover layer sample has been determined with an HPGe gamma-spectrometer. The gamma dose rate was estimated from the measured radionuclide concentrations and the results were compared with the measured ambient dose equivalent rate. Based on the measured results before and after covering, the ambient dose equivalent rates were 76 (67-85) nSv/h before and 86 (83-89) nSv/h after covering, radon exhalation was 9 (6-12) mBq/m 2 s before and 14 (5-28) mBq/m 2 s after covering, the outdoor radon concentrations were 11 (9-16) before and 13 (10-22) Bq/m 3 after covering and the soil gas radon concentrations were 6 (3-8) before and 24 (14-40) kBq/m 3 after covering. Soil gas permeability measurements were 1E-11 (7E-12-1E-11) and 1E-12 (5E-13-1E-12) m 2 and the calculated geogenic radon potential values were 6 (3-8) and 12 (6-21) before and after the covering. The main radionuclide concentrations of the drilling mud were C U-238 12 (10-15) Bq/kg, C Ra-226 31 (18-40) Bq/kg, C Th-232 35 (33-39) Bq/kg and C K-40 502 (356-673) Bq/kg. The same radionuclide concentrations in the clay were C U-238 31 (29-34) Bq/kg, C Ra-226 45 (40-51) Bq/kg, C Th-232 58 (55

  5. Local system for control by console-mobile crane for russian depository of fissionable materials

    International Nuclear Information System (INIS)

    Troshchenko, V.G.; Kapustin, V.N.; Zinina, N.V.; Derbyshev, S.A.

    2005-01-01

    Description of crane of console-mobile type used for transportation of fissionable materials in depository with local control system is represented. Local control system realizes program control in real time [ru

  6. UK modelling of thermal effects on leakage from hard rock depositories

    International Nuclear Information System (INIS)

    Bourke, P.J.

    1980-01-01

    Thermally induced stress through and around depositories have been calculated assuming the rock to have constant mechanical properties obtained from laboratory measurements and ignoring the effects of existing fractures. After allowing for probable values of the natural stress field, regions of net tension and high shear stress which might cause new fractures were found. This analysis is, however, not yet considered to be reliable because of uncertainty about the above assumptions. Further, even if it is accurate, it is incomplete because it is still not possible to relate quantitatively calculated stresses to changes in permeability and porosity due to changes in existing fractures or initiation of new ones. Accordingly, further theoretical work is being done to plan an underground study of the effects of heating on a well defined fracture. Measurements of strain and modulus will be made to investigate the validity of the mechanical assumptions and hydraulic data will be obtained to relate stress to resistance to flow. It is hoped that further analysis will then allow an assessment of the importance of thermal stress around a depository to be made

  7. Transition to Electronic Access of Government Information: Are the Depository Libraries Prepared?

    Science.gov (United States)

    Vaughan, Liwen Qiu; Dolan, Elizabeth

    1998-01-01

    Examines the readiness of depository libraries in Canada to adopt new technologies for disseminating government publications. Findings are reported on current use of different publication formats, type of help sought by users, staff skills and training needs, adequacy of physical and financial resources, support from governing bodies, and…

  8. Study of basic biopolymer as proton membrane for fuel cell systems

    International Nuclear Information System (INIS)

    Ramirez-Salgado, Joel

    2007-01-01

    Up to now, many research groups work to improve the electrical and mechanical properties of membranes with a low cost of production. The biopolymers could be an answer to produce proton membranes at low cost. This work demonstrates that the intrinsic membrane polymer and clays properties can help to develop a novel proton exchange membranes. Biopolymer composites (chitosan-oxide compounds) present conductivity between 10 -3 and 10 -2 S cm -1 . The measurements were calculated by EIS (1 MHz-0.05 Hz) using the two-electrode configuration. Different oxides were used: MgO, CaO, SiO 2 , Al 2 O 3 . The ionic conductivities were compared with Nafion (registered)'s in the same conditions of P and T. The catalyst layer/membrane ensemble was made during the design with the subsequent demonstration as membrane electrode assemblies and finally the fuel cell was built. Our focus was to increase the compatibility between the proton basic polymer exchange membrane and basic clays as CaO and test a new kind of fuel cell

  9. 77 FR 12897 - Self-Regulatory Organizations; Pacific Securities Depository Trust Company; Order Cancelling...

    Science.gov (United States)

    2012-03-02

    ... SECURITIES AND EXCHANGE COMMISSION [Release No. 34-66460; File No. 600-10] Self-Regulatory Organizations; Pacific Securities Depository Trust Company; Order Cancelling Clearing Agency Registration...(a)(3) of the Act \\14\\ provides that in the event any self-regulatory organization is no longer in...

  10. Fuel performance-REP, Seminars on nuclear fuel performance based on basic underlining phenomena, proceedings

    International Nuclear Information System (INIS)

    2008-01-01

    Description: The need for further improving the understanding of basic phenomena underlying nuclear fuel behaviour has been recognised both by fuel vendors, experts in fuel research in the different laboratories and committees and working groups coordinating international activities. The OECD/NEA Nuclear Science Committee has established an Experts Group addressing this issue. This has led to establishing an International Fuel Performance Experiments Database (IFPE) that should help model evaluation and validation. Many years ago the IAEA established an International Working Group on Fuel Performance and Technology (IWGFPT) that led to the FUMEX-I and FUMEX-II (Fuel Modelling Exercise) which has had an important impact on code improvements. Both international organisations, with the support of national organisations, co-operate in establishing and maintaining the Database and to build confidence in the predictive power of the models through international comparison exercises. But above all the different parties have agreed that seminars focussed on specific phenomena would be beneficial to exchange current knowledge, identify outstanding problems and agree on common action that would lead to improved understanding of the phenomena. A series of three seminars has been initiated by the Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), Framatome and Cogema under the aegis of the OECD/NEA and the IAEA. 1. Thermal Performance of High Burn-Up LWR Fuel at Cadarache, France, from 3 to 6 of March 1998. Thermal performance occupies the most important aspect of the fuel performance modelling. Not only is it extremely important from a safety point of view, but also many of the material properties of interest and behaviour, such as transport properties like fuel creep and fission gas release are thermally activated processes. Thus, in order to model these processes correctly, it is critical to calculate temperatures and their distribution as accurately as

  11. The Federal Depository Library Program (FDLP), Academic Libraries, and Access to Government Information

    Science.gov (United States)

    Jaeger, Paul T.; Bertot, John Carlo; Shuler, John A.

    2010-01-01

    The electronic environment has significantly shifted library capabilities and user expectations for the delivery of government information and services. At the same time, many laws of the federal government have pushed for the creation and distribution of government information through electronic channels. However, the Federal Depository Library…

  12. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    Nagano, M.; Sakurai, S.; Yamaguchi, H.

    1997-01-01

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  13. 5 CFR 3201.103 - Prohibition on acquisition, ownership, or control of securities of FDIC-insured depository...

    Science.gov (United States)

    2010-01-01

    ... described in paragraph (a) of this section if: (i) The security was acquired by inheritance, gift, stock...-insured depository institution as custodian or trustee of accounts containing tax-deferred retirement...

  14. Investigation and basic evaluation for ultra-high burnup fuel cladding material

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi

    2001-03-01

    In ultra-high burnup of the power reactor, it is an essential problem to develop the cladding with excellent durability. First, development history and approach of the safety assessment of Zircaloy for the high burnup fuel were summarized in the report. Second, the basic evaluation and investigation were carried out on the material with high practicability in order to select the candidate materials for the ultra-high burnup fuel. In addition, the basic research on modification technology of the cladding surface was carried out from the viewpoint of the addition of safety margin as a cladding. From the development history of the zirconium alloy including the Zircaloy, it is hard to estimate the results of in-pile test from those of the conventional corrosion test (out-pile test). Therefore, the development of the new testing technology that can simulate the actual environment and the elucidation of the corrosion-controlling factor of the cladding are desired. In cases of RIA (Reactivity Initiated Accident) and LOCA (Loss of Coolant Accident), it seems that the loss of ductility in zirconium alloys under heavy irradiation and boiling of high temperature water restricts the extension of fuel burnup. From preliminary evaluation on the high corrosion-resistance materials (austenitic stainless steel, iron or nickel base superalloys, titanium alloy, niobium alloy, vanadium alloy and ferritic stainless steel), stabilized austenitic stainless steels with a capability of future improvement and high-purity niobium alloys with a expectation of the good corrosion resistance were selected as candidate materials of ultra-high burnup cladding. (author)

  15. Design and Implementation of Corporate Actions on a Decentralized Securities Depository

    OpenAIRE

    Hedin, Jonas

    2017-01-01

    Trading securities is a process that requires multiple trusted intermediaries to ensure that the trade is done correctly. The securities industry is therefore very slow and expensive; the central securities depository (CSD) being one of the main contributors to the disruption. In an effort to fix this, financial institutions has recently started looking into the blockchain technology; the innovation behind the cryptocurrency Bitcoin. Bitcoin is a digital currency that can be traded peer-to-pe...

  16. Basic concept of fuel safety design and assessment for sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Nakae, Nobuo; Baba, Toshikazu; Kamimura, Katsuichiro

    2013-03-01

    'Philosophy in Safety Evaluation of Fast Breeder Reactors' was published as a guideline for safety design and safety evaluation of Sodium-Cooled Fast Reactor in Japan. This guideline points out that cladding creep and swelling due to internal pressure should be taken into account since the fuel is used under high temperature and high burnup, and that fuel assembly deformation and the prevention from coolant channel blockage should be taken into account in viewpoints of nuclear and thermal hydraulic design. However, the requirements including their criteria and evaluation items are not described. Two other domestic guidelines related to core design are applied for fuel design of fast reactor, but the description is considered to not be enough to practically use. In addition, technical standard for nuclear fuel used in power reactors is also applied for fuel inspection. Therefore, the technical standard and guideline for fuel design and safety evaluation are considered to be very important issue for nuclear safety regulation. This document has been developed according to the following steps: The guidelines and the technical standards, which are prepared in foreign countries and international organization, were reviewed. The technical background concerning fuel design and safety evaluation for fast reactor was collected and summarized in the world wide scale. The basic concept of fuel safety design and assessment for sodium-cooled fast reactor was developed by considering a wide range of views of the specialists in Japan. In order to discuss the content with foreign specialists IAEA Consultancy Meetings have been held on January, 2011 and January, 2012. The participants of the meeting came from USA, UK, EC, India, China and South Korea. The specialists of IAEA and JNES were also joined. Although this document is prepared for application to 'Monju'(prototype LMFR), it may be applied to experimental, demonstration and commercial types of LMFR after revising it by taking

  17. 77 FR 3531 - Self-Regulatory Organizations; The Depository Trust Company; Order Approving Proposed Rule Change...

    Science.gov (United States)

    2012-01-24

    ... Organizations; The Depository Trust Company; Order Approving Proposed Rule Change To Enhance Risk Management.... Description The rule change will enhance the risk management controls associated with DTC's Receiver... Participant's net debit cap is placed on a pending (recycling) queue until another transaction creates...

  18. Mykobiota of the air of depositories and documents of V. I. Vernadsky National Library of Ukraine (Historical aspect, the research since 1992

    Directory of Open Access Journals (Sweden)

    Subbota A. H.

    2017-01-01

    Full Text Available The results of the first in Ukraine systematic monitoring research of the mycological state of the air of depositories and documents of V. I. Vernadsky National Library of Ukraine are retrospectively generalized and presented. Since 1992, in a comparative aspect, the species composition of the mycobiota of air has been studied in its quiescent state and in the process of sanitary-hygienic processing of fonds, as well as the seasonal dynamics of micromycetes. About 15,000 strains of microscopic fungi (micromycetes belonging to 81 species of 26 genera, 3 departments of Zygomycotina, Ascomycotina i Mytosporic fungi (Deuteromycotina were isolated and identified from the air of depositories and documents. The dependence of the quantitative and taxonomic characteristics of the mycobiota of air on the storage conditions and the physical state of the documents was determined, and the mycological index of the ecological state of the storages and documents was clarified. The destructive properties of micromycetes isolated from mycobiota of the depositories have been studied.

  19. 17 CFR 270.17f-4 - Custody of investment company assets with a securities depository.

    Science.gov (United States)

    2010-04-01

    ... available concerning the internal accounting controls and financial strength of the custodian; and (3... controls and financial strength of the securities depository; and (2) The fund has implemented internal... 17 Commodity and Securities Exchanges 3 2010-04-01 2010-04-01 false Custody of investment company...

  20. Basic research on cermet nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ohashi, Hiroshi; Sto, Seichi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering; Takano, Masahide; Minato, Kazuo; Fukuda, Kosaku

    1998-01-01

    Production of cermet nuclear fuel having fine uranium dioxide (UO{sub 2}) particles dispersed in matrix metal requires basic property data on the compatibility of matrix metal with fission product compounds. It is thermodynamically suggested that, as burnup increases, cesium in oxide fuel reacts with the fuel, other fission products or cladding pipe and produces cesium uranates, cesium molybdate, or cesium chromate in stainless steel cladding pipe. Attempt was made to measure the thermal expansion coefficient and thermal conductivity of cesium uranates (Cs{sub 2}UO{sub 4} and Cs{sub 2}U{sub 2}O{sub 7}), cesium molybdate (Cs{sub 2}MoO{sub 4}) and cesium chromate (Cs{sub 2}CrO{sub 4}). Thermal expansion was measured by X-ray diffraction and determined by Cohen`s method. Thermal conductivity was obtained by measuring thermal diffusion by laser flash method. The thermal expansion of Cs{sub 2}UO{sub 4} and Cs{sub 2}U{sub 2}O{sub 7} is as low as 1.2% for the former and 1.0% for the latter, up to 1000K. The thermal expansion of Cs{sub 2}MoO{sub 4} is as high as that of Cs{sub 2}CrO{sub 4}, 2.1% for the former and 2.5% for the latter at temperatures from room temperature to 873K. Average thermal expansion in this temperature range is 4.4 x 10{sup -5} K{sup -1} for Cs{sub 2}MoO{sub 4} and 4.2 x 10{sup -5} K{sup -1}. The thermal expansion of Cs{sub 2}CrO{sub 4} is four times higher than that of UO{sub 2} and five times higher than that of Cr{sub 2}O{sub 3}. The thermal conductivity of Cs{sub 2}UO{sub 4} is nearly equal to that of Cs{sub 2}U{sub 2}O{sub 7} in absolute value and temperature dependency. Cs{sub 2}U{sub 2}O{sub 7}, having different thermal conductivity between {alpha} and {beta} phases, shows higher conductivity with {beta} than with {alpha}, about 1/4 of that of UO{sub 2} at 1000K. The thermal conductivity of Cs{sub 2}CrO{sub 4} is nearly equal to that of Cs{sub 2}MoO{sub 4} in absolute value and temperature dependency. (N.H.)

  1. Basic design study on plutonium electro-refining facility of oxide fuel pyroelectrochemical reprocessing

    International Nuclear Information System (INIS)

    Ogura, Kenji; Kondo, Naruhito; Kamoshida, Hiroshi; Omori, Takashi

    2001-02-01

    The test facility basic design, utility necessity and estimation cost of the Oxide Fuel Pyro-process for the use of Chemical Processing Facility (CPF) of JNC have been studied with the information of the previous year concept study and the additional conditions. Drastic down sizing design change or the building reconstruction is necessary to place the Oxide Fuel Pyro-process Facility in the laboratory ''C'', because it is not possible to reserve enough maintenance space and the weight of the facility is over the acceptable limit of the building. A further study such as facility down sizing, apparatus detail design and experiment detail process treatment has to be planned. (author)

  2. Fuel cell technology for classroom instruction. Basic principles, experiments, work sheets. 2. ed.

    Energy Technology Data Exchange (ETDEWEB)

    Voigt, Cornelia; Hoeller, Stefan; Kueter, Uwe

    2009-07-01

    This book provides a clear introduction and overview to fuel cell technology and its associated subject areas. Examples of experiments using solar cells, electrolysis and fuel cells convey the knowledge for forthcoming tests in an understandable manner. The preparation of classroom experiments is made considerably easier for the teacher thanks to the experiment work sheets. These contain the necessary information concerning the material, set-up and execution of the experiment, and questions for evaluation purposes. Online-Shop The training documents and student work sheets combine the basic knowledge, questions and answers, and are ideal for copying. A comprehensive glossary at the end of the book explains all the important technical terms. (orig.)

  3. Present state of the design and realization of regional radioactive waste depositories and waste acceptance criteria for disposal. Soucasny stav reseni a realizace regionalnich ulozist RA odpadu a kriteria prijatelnosti techto odpadu k ukladani

    Energy Technology Data Exchange (ETDEWEB)

    Kortus, J [Chemoprojekt, Prague (Czechoslovakia)

    1988-06-01

    Surface type regional depositories for radioactive wastes from nuclear power plants are described in detail. The depository of the Mochovce nuclear power plant is located near the plant, that of the Dukovany nuclear power plant is directly on the premises of the plant. Particular attention is paid to the design of the monolithic reinforced concrete pits, draining of rainwater from their surface, draining of seeping rainwater from the pit environment by means of a double drainage system, and insulation of the pits against water. The construction of the Mochovce depository started in 1987; some experience gained from this activity is presented. The radioactive waste acceptance criteria for depositories of this kind, based on safety analysis, are given. (author). 2 figs.

  4. Basic data biogas Germany. Solid fuels, biofuels, biogas; Basisdaten Bioenergie Deutschland. Festbrennstoffe, Biokraftstoffe, Biogas

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-09-15

    The brochure ''Basic data biogas Germany'' gives statistical information about (a) renewable energies: primary energy consumption, power generation, energy supply, avoidance of greenhouse gases; (b) Solid fuels: energetic utilization, wood pellets, energy consumption, comparison to heating oil; (c) Biofuels: consumption, bioethanol, biodiesel, vegetable oils; (d) Biogas: biogas power plants, energy content, production, legal aspects.

  5. Basic data biogas Germany. Solid fuels, biofuels, biogas; Basisdaten Bioenergie Deutschland. Festbrennstoffe - Biokraftstoffe - Biogas

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-08-15

    The brochure ''Basic data biogas Germany'' gives statistical information about (a) renewable energies: primary energy consumption, power generation, energy supply, avoidance of greenhouse gases; (b) Solid fuels: energetic utilization, wood pellets, energy consumption, comparison to heating oil; (c) Biofuels: consumption, bioethanol, biodiesel, vegetable oils; (d) Biogas: biogas power plants, energy content, production, legal aspects.

  6. 78 FR 13917 - Self-Regulatory Organizations; The Depository Trust Company; Notice of Filing Amendment No. 1 and...

    Science.gov (United States)

    2013-03-01

    ... financial stability by, among other things, promoting uniform risk management standards for systemically.... Introduction On December 28, 2012, The Depository Trust Company (``DTC'') filed with the Securities and... systemic risk since the reversals may override DTC's risk management controls such as the Collateral...

  7. 12 CFR 204.135 - Shifting funds between depository institutions to make use of the low reserve tranche.

    Science.gov (United States)

    2010-01-01

    ... SYSTEM BOARD OF GOVERNORS OF THE FEDERAL RESERVE SYSTEM RESERVE REQUIREMENTS OF DEPOSITORY INSTITUTIONS... prevent evasions of the requirements of that section. Section 19(b)(2) establishes general reserve... that is exempt from transaction account reserve requirements. [57 FR 38429, Aug. 25, 1992] ...

  8. 78 FR 9762 - Self-Regulatory Organizations; The Depository Trust Company; Notice of Designation of a Longer...

    Science.gov (United States)

    2013-02-11

    ... Presentments and Issuances of Money Market Instruments February 5, 2013. On December 17, 2012, The Depository... processing of maturity and income presentments and issuances of money market instruments. The Commission... Division of Trading and Markets, pursuant to delegated authority.\\7\\ \\7\\ 17 CFR 200.30-3(a)(31). Kevin M. O...

  9. 12 CFR 204.128 - Deposits at foreign branches guaranteed by domestic office of a depository institution.

    Science.gov (United States)

    2010-01-01

    ... SYSTEM BOARD OF GOVERNORS OF THE FEDERAL RESERVE SYSTEM RESERVE REQUIREMENTS OF DEPOSITORY INSTITUTIONS.... (b) Section 19 of the Federal Reserve Act which establishes reserve requirements does not apply to... in 1918 that the requirements of section 19 as to reserves to be carried by member banks do not apply...

  10. Basic principles

    International Nuclear Information System (INIS)

    Wilson, P.D.

    1996-01-01

    Some basic explanations are given of the principles underlying the nuclear fuel cycle, starting with the physics of atomic and nuclear structure and continuing with nuclear energy and reactors, fuel and waste management and finally a discussion of economics and the future. An important aspect of the fuel cycle concerns the possibility of ''closing the back end'' i.e. reprocessing the waste or unused fuel in order to re-use it in reactors of various kinds. The alternative, the ''oncethrough'' cycle, discards the discharged fuel completely. An interim measure involves the prolonged storage of highly radioactive waste fuel. (UK)

  11. 12 CFR 204.131 - Participation by a depository institution in the secondary market for its own time deposits.

    Science.gov (United States)

    2010-01-01

    ... RESERVE SYSTEM BOARD OF GOVERNORS OF THE FEDERAL RESERVE SYSTEM RESERVE REQUIREMENTS OF DEPOSITORY... designed to distinguish between time deposits and demand deposits for federal reserve requirement purposes... also used to distinguish between types of deposits for reserve requirement purposes. Effective April 1...

  12. Implications of Harmonizing the Future of the Federal Depository Library Program within E-Government Principles and Policies

    Science.gov (United States)

    Shuler, John A.; Jaeger, Paul T.; Bertot, John Carlo

    2010-01-01

    For more than 150 years, the United States Government Printing Office (GPO), along with its Federal Depository Library Program (FDLP), has supported an informed citizenry and democracy by ensuring access and preservation to a broad swath of federal government information. This collaborative national public information program between local…

  13. 77 FR 3075 - Resolution Plans Required for Insured Depository Institutions With $50 Billion or More in Total...

    Science.gov (United States)

    2012-01-23

    ... business day of the institution's failure (two business days if the failure occurs on a day other than... the continuing exposure of the banking industry to the risks of insolvency of large and complex insured depository institutions, an exposure that can be mitigated with proper resolution planning. The...

  14. Status of the natural and enriched uranium market: the basic economical factor for the development of the fuel cycle

    International Nuclear Information System (INIS)

    Nochev, T.

    1999-01-01

    Status of the Natural and Enriched Uranium Market - the Basic. Economical Factor for the Development of the Fuel Cycle An overview of the status of the natural and enriched uranium market has been performed and it offers a possibility to estimate the changes and tendencies, the knowledge of which is needed in negotiations about the fresh fuel. The simplified financial analysis presented here demonstrates the economical profitability of the storage of the spent fuel making now the allocations for the future reprocessing

  15. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    International Nuclear Information System (INIS)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

  16. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

  17. Ethanol Basics

    Energy Technology Data Exchange (ETDEWEB)

    None

    2015-01-30

    Ethanol is a widely-used, domestically-produced renewable fuel made from corn and other plant materials. More than 96% of gasoline sold in the United States contains ethanol. Learn more about this alternative fuel in the Ethanol Basics Fact Sheet, produced by the U.S. Department of Energy's Clean Cities program.

  18. The U.S. Depository Library Program and the Separation of Powers: A Response to the Librarians' Manifesto.

    Science.gov (United States)

    Sprehe, J. Timothy; Morton, Bruce

    1993-01-01

    Two articles respond to the "Librarians' Manifesto," which discusses the U.S. Depository Library Program (DLP). The first urges that DLP be legislatively removed from the Government Printing Office. The second views the manifesto as the mindset of academic librarians, who think the DLP is a collection of entitlement programs for academic…

  19. Initial assessment of the thermal stresses around a radioactive waste depository in hard rock

    International Nuclear Information System (INIS)

    Hodgkinson, D.P.; Bourke, P.J.

    1980-01-01

    The disposal of heat emitting radioactive waste into hard rock should result in temperature rises and thermal gradients over distances of several hundred metres for several centuries. The consequent constrained thermal expansion of the rock would induce stresses which have important implications for possible water-borne leakage of radionuclides and for depository design. These problems are assessed by considering a simplified mathematical model for which analytic solutions to the temperature and stress fields are derived. (author)

  20. Revision of the second basic plans of power reactor development in Power Reactor and Nuclear Fuel Development Corporation

    International Nuclear Information System (INIS)

    1978-01-01

    Revision of the second basic plans concerning power reactor development in PNC (Power Reactor and Nuclear Fuel Development Corporation) is presented. (1) Fast breeder reactors: As for the experimental fast breeder reactor, after reaching the criticality, the power is raised to 50 MW thermal output within fiscal 1978. The prototype fast breeder reactor is intended for the electric output of 200 MW -- 300 MW, using mixed plutonium/uranium oxide fuel. Along the above lines, research and development will be carried out on reactor physics, sodium technology, machinery and parts, nuclear fuel, etc. (2) Advanced thermal reactor: The prototype advanced thermal reactor, with initial fuel primarily of slightly enriched uranium and heavy water moderation and boiling water cooling, of 165 MW electric output, is brought to its normal operation by the end of fiscal 1978. Along the above lines, research and development will be carried out on reactor physics, machinery and parts, nuclear fuel, etc. (Mori, K

  1. International Depository Authority and its Role in Microorganism's Deposition.

    Science.gov (United States)

    Parashar, Abhishek

    2017-08-01

    After the World Trade Organization's agreement on Trade Related Aspect of Intellectual Property Right, patents have come into major play. Patenting of work related to live organisms that have medical, agricultural and other uses is always a tedious, complicated and controversial job. In view of this an agreement called as Budapest Treaty was passed in 1977 for deposition of microorganisms in culture collection centers for patent purpose. To make a culture collection center an IDA the culture center has to follow rules and regulations made in Budapest Treaty. Today several culture collection centers are working as International Depository Authority (IDA) in India and abroad that are storing microorganisms for patent purpose. India has two IDA units; one is Microbial Culture Collection in National Centre for Cell Science Pune, India, and second one is Microbial Type Culture Collection and Gene Bank at Institute of Microbial Technology Chandigarh, India. IDA units are important for valuable sample depositions, supply of cultures, characterization of samples and conservation of biological materials.

  2. Alternative Fuels Data Center: Ethanol Fueling Stations

    Science.gov (United States)

    ... More in this section... Ethanol Basics Benefits & Considerations Stations Locations Infrastructure fueling stations by location or along a route. Infrastructure Development Learn about ethanol fueling infrastructure; codes, standards, and safety; and ethanol equipment options. Maps & Data E85 Fueling Station

  3. Fuel performance and operation experience of WWER-440 fuel in improved fuel cycle

    International Nuclear Information System (INIS)

    Gagarinski, A.; Proselkov, V.; Semchenkov, Yu.

    2007-01-01

    The paper summarizes WWER-440 second-generation fuel operation experience in improved fuel cycles using the example of Kola NPP units 3 and 4. Basic parameters of fuel assemblies, fuel rods and uranium-gadolinium fuel rods, as well as the principal neutronic parameters and burn-up achieved in fuel assemblies are presented. The paper also contains some data concerning the activity of coolant during operation (Authors)

  4. Basic principles of forest fuel reduction treatments

    Science.gov (United States)

    James K. Agee; Carl N. Skinner

    2005-01-01

    Successful fire exclusion in the 20th century has created severe fire problems across the West. Not every forest is at risk of uncharacteristically severe wildfire, but drier forests are in need of active management to mitigate fire hazard. We summarize a set of simple principles important to address in fuel reduction treatments: reduction of surface fuels, increasing...

  5. Alternative fuels in fire debris analysis: biodiesel basics.

    Science.gov (United States)

    Stauffer, Eric; Byron, Doug

    2007-03-01

    Alternative fuels are becoming more prominent on the market today and, soon, fire debris analysts will start seeing them in liquid samples or in fire debris samples. Biodiesel fuel is one of the most common alternative fuels and is now readily available in many parts of the United States and around the world. This article introduces biodiesel to fire debris analysts. Biodiesel fuel is manufactured from vegetable oils and/or animal oils/fats. It is composed of fatty acid methyl esters (FAMEs) and is sold pure or as a blend with diesel fuel. When present in fire debris samples, it is recommended to extract the debris using passive headspace concentration on activated charcoal, possibly followed by a solvent extraction. The gas chromatographic analysis of the extract is first carried out with the same program as for regular ignitable liquid residues, and second with a program adapted to the analysis of FAMEs.

  6. The development of basic glass formulations for solidifying HLW from nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Jiang Yaozhong; Tang Baolong; Zhang Baoshan; Zhou Hui

    1995-01-01

    Basic glass formulations 90U/19, 90U/20, 90Nd/7 and 90Nd/10 applied in electric melting process are developed by using the mathematical model of the viscosity and electric resistance of waste glass. The yellow phase does not occur for basic glass formulations 90U/19 and 90U/20 solidifying HLW from nuclear fuel reprocessing plant when the waste loading is 20%. Under the waste loading is 16%, the process and product properties of glass 90U/19 and 90U/20 come up to or surpass the properties of the same kind of foreign waste glasses, and other properties are about the same to them of foreign waste glasses. The process and product properties of basic glass formulations 90Nd/7 and 90Nd/10 used for the solidification of 'U replaced by Nd' liquid waste are almost similar to them of 90U/19 and 90U/20. These properties fairly meet the requirements of 'joint test' (performed at KfK-INE, Germany). Among these formulations, 90Nd/7 is applied in cold engineering scale electric melting test performed at KfK-INE in Germany. The main process properties of cold test is similar to laboratory results

  7. Depository of ampoule ionizing radiation sources on the basis of stand complex Baikal-I. Operation experience and application perspectives

    International Nuclear Information System (INIS)

    Ganzha, V.V.; Boltovskij, S.A.; Kolbaenkov, A.N.; Meshin, M.M.; Nasonov, S.G.; Pivovarov, O.S.; Storozhenko, A.S.; Yakovlev, V.V.

    2001-01-01

    Depository of ampoule sources of ionizing radiation (ASIR) on the basis of stand complex Baikal-I was founded an put into operation in 1995. It is intended for prolonged storage of the spent ASIR from a different institutions of Kazakhstan. To the present time a more than 10000 spent ASIR with activity more than 2000 Ci were taken and placed for storage

  8. Ethanol Basics (Fact Sheet)

    Energy Technology Data Exchange (ETDEWEB)

    2015-01-01

    Ethanol is a widely-used, domestically-produced renewable fuel made from corn and other plant materials. More than 96% of gasoline sold in the United States contains ethanol. Learn more about this alternative fuel in the Ethanol Basics Fact Sheet, produced by the U.S. Department of Energy's Clean Cities program.

  9. Fuel cell electrolyte membrane with basic polymer

    Science.gov (United States)

    Larson, James M.; Pham, Phat T.; Frey, Matthew H.; Hamrock, Steven J.; Haugen, Gregory M.; Lamanna, William M.

    2012-12-04

    The present invention is an electrolyte membrane comprising an acid and a basic polymer, where the acid is a low-volatile acid that is fluorinated and is either oligomeric or non-polymeric, and where the basic polymer is protonated by the acid and is stable to hydrolysis.

  10. Criticality safety analyses in SKODA JS a.s

    International Nuclear Information System (INIS)

    Mikolas, P.; Svarny, J.

    1999-01-01

    This paper describes criticality safety analyses of spent fuel systems for storage and transport of spent fuel performed in SKODA JS s.r.o.. Analyses were performed for different systems both at NPP site including originally designed spent fuel pool with a large pitch between assemblies without any special absorbing material, high density spent fuel pool with an additional absorption by boron steel, depository rack for fresh fuel assemblies with a very large pitch between fuel assemblies, a container for transport of fresh fuel into the reactor pool and a cask for transport and storage of spent fuel and container for final storage depository. required subcriticality has been proven taking into account all possible unfavourable conditions, uncertainties etc. In two cases, burnup credit methodology is expected to be used. (Authors)

  11. On possibility of degradation of lava-like fuel-containing materials of the 4-th block of Chernobyl NPP under internal self-irradiation by alpha-particle sources

    International Nuclear Information System (INIS)

    Pazukhin, Eh. M.; Borovoj, A.A.; Rudya, K.G.

    2002-01-01

    It is shown that internal self-irradiation by alpha-particle beam cannot be a cause of change of strength characteristics of silicate matrix and so a cause of degradation of Chernobyl lava-like materials. A new method is proposed for management with lava-like fuel-containing materials of the 4-th block: vitrification in smelter unit situated in bubbler-basin and storage of prepared immobilized compacts in corresponding depositories [ru

  12. Fuel cells:

    DEFF Research Database (Denmark)

    Sørensen, Bent

    2013-01-01

    A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil and nucl......A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil...... and nuclear fuel-based energy technologies....

  13. Modeling, simulation and analysis of a securities settlement system:The case of Central Securities Depository of Mexico

    OpenAIRE

    Muñoz, David F; Palacios, Arturo; Lascurain, Miguel

    2012-01-01

    The Instituto para el Depósito de Valores (INDEVAL) is the Central Securities Depository of Mexico. It is the only Mexican institution authorized to perform, in an integrated manner, the activities of safe-keeping, custody, management, clearing, settlement and transfer of securities. In this article, we report the modeling, simulation and analysis of a new Securities Settlement System (SSS) implemented by INDEVAL, as part of a project for the implementation of a safer and more efficient opera...

  14. Modeling, simulation and analysis of a securities settlement system: the case of Central Securities Depository of Mexico

    OpenAIRE

    Muñoz, David F.; Palacios, Arturo; de Lascurain, Miguel

    2012-01-01

    The Instituto para el Depósito de Valores (INDEVAL) is the Central Securities Depository of Mexico. It is the only Mexican institution authorized to perform, in an integrated manner, the activities of safe-keeping, custody, management, clearing, settlement and transfer of securities. In this article, we report the modeling, simulation and analysis of a new Securities Settlement System (SSS) implemented by INDEVAL, as part of a project for the implementation of a safer and more efficient opera...

  15. Ammonia as a Suitable Fuel for Fuel Cells

    International Nuclear Information System (INIS)

    Lan, Rong; Tao, Shanwen

    2014-01-01

    Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5 wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel cells are briefly reviewed.

  16. Multi-fuel reformers for fuel cells used in transportation. Phase 1: Multi-fuel reformers

    Science.gov (United States)

    1994-05-01

    DOE has established the goal, through the Fuel Cells in Transportation Program, of fostering the rapid development and commercialization of fuel cells as economic competitors for the internal combustion engine. Central to this goal is a safe feasible means of supplying hydrogen of the required purity to the vehicular fuel cell system. Two basic strategies are being considered: (1) on-board fuel processing whereby alternative fuels such as methanol, ethanol or natural gas stored on the vehicle undergo reformation and subsequent processing to produce hydrogen, and (2) on-board storage of pure hydrogen provided by stationary fuel processing plants. This report analyzes fuel processor technologies, types of fuel and fuel cell options for on-board reformation. As the Phase 1 of a multi-phased program to develop a prototype multi-fuel reformer system for a fuel cell powered vehicle, the objective of this program was to evaluate the feasibility of a multi-fuel reformer concept and to select a reforming technology for further development in the Phase 2 program, with the ultimate goal of integration with a DOE-designated fuel cell and vehicle configuration. The basic reformer processes examined in this study included catalytic steam reforming (SR), non-catalytic partial oxidation (POX) and catalytic partial oxidation (also known as Autothermal Reforming, or ATR). Fuels under consideration in this study included methanol, ethanol, and natural gas. A systematic evaluation of reforming technologies, fuels, and transportation fuel cell applications was conducted for the purpose of selecting a suitable multi-fuel processor for further development and demonstration in a transportation application.

  17. Survey of a technology to introduce the waste-fueled power generation. Basic manual for introduction of the waste-fueled power generation; Haikibutsu hatsuden donyu gijutsu chosa. Haikibutsu hatsuden donyu kihon manual

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-01

    Local government offices, etc., which are expected to shoulder responsibility for introducing the waste-fueled power generation, want to need exact information on technical information concerning the waste-fueled power generation and the method to materialize the introduction plan, etc. Therefore, Electric Power Development Co. surveyed and studied it under the contract with NEDO. The results were collected together as a basic manual for introduction of the waste-fueled power generation. As an outline of the waste-fueled power generation, the manual explains the significance, the present situation and potentials, the waste-fueled power system, an outline of working out the waste-fueled power generation plan, an outline of construction and operation/maintenance of the waste-fueled power generation, an outline of various systems relating to the waste-fueled power generation, etc. As the items for the study of making a concrete plan for power generation equipment, the manual explains the amount of refuse to be incinerated, the present status of generation capacity as viewed from the quality of refuse, the quality of refuse and the design of power generation equipment, boiler efficiency, power generation efficiency, construction cost and operation cost, etc. In addition, the paper describes a case study of the waste-fueled power generation plan. 118 figs., 39 tabs.

  18. Basic study on cermet fuels

    Energy Technology Data Exchange (ETDEWEB)

    Takano, Kimihide; Sato, Seichi; Ohashi, Hiroshi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering; Minato, Kazuo; Fukuda, Kosaku

    1996-01-01

    Cesium is a major nuclear fission product which is volatile and corrosive and it is able to interact with fuels, cladding tubes and/or other fission products resulting in productions of various compounds. The presence of those compounds may give rise to physical and chemical interactions between fuels and the cladding tube, resulting in changes in their heat transfer coefficients. In this study, some cesium uranates were prepared in the laboratory. Then, Cs{sub 2}UO{sub 4}, Cs{sub 2}U{sub 2}O{sub 7} and Cs{sub 2}U{sub 4}O{sub 12} were thermodynamically analyzed by phase equilibrium calculation using the calculation code, CHEMSAGE. And physicochemical properties of these compounds were investigated. The reaction conditions for producing the three compounds were determined. The equilibrium diagram revealed that Cs{sub 2}UO{sub 4} is stable in a wider range for the partial pressures of oxygen and cesium than Cs{sub 2}U{sub 2}O{sub 7} or Cs{sub 2}U{sub 4}O{sub 12}. Some orange colored product was obtained from the reaction of Cs{sub 2}CO{sub 3} and U{sub 3}O{sub 8} in an electric furnace and Cs{sub 2}U{sub 2}O{sub 7} but not Cs{sub 2}UO{sub 4} was identified by X-ray diffraction of the product. (M.N.)

  19. Ammonia as a suitable fuel for fuel cells

    Directory of Open Access Journals (Sweden)

    Rong eLan

    2014-08-01

    Full Text Available Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel cells are briefly reviewed.

  20. A study on the basic CVD process technology for TRISO coated particle fuel

    International Nuclear Information System (INIS)

    Choi, D. J.; Cheon, J. H.; Keum, I. S.; Lee, H. S.; Kim, J. G.

    2006-03-01

    Hydrogen energy has many advantages and is suitable as alternative energy of fossil fuel. The study of nuclear hydrogen production has performed at present. For nuclear hydrogen production, it is needed the study of VHTR(Very High Temperature Reactor) and TRISO(TRI-iSOtropic) coated fuel. TRISO coated fuel particle deposited by FBCVD(Fludized Bed CVD) method is composed of three isotropic layers: Inner Pyrolytic Carbon (IPyC), Silicon Carbide (SiC), Outer Pyrolytic Carbon (OPyC) layers. Silicon carbide was chemically vapor deposed on graphite substrate using methyltrichlorosilane (CH 3 SiCl 3 ) as a source in hydrogen atmosphere. The effect of deposition temperature and input gas ratios ( α=Q H2 /Q MTS =P H2 /P MTS ) was investigated in order to find out characteristics of silicon carbide layer. From results of those, SiC-TRISO coating deposition was conducted and achieved. Zirconium carbide layer as an advanced material of silicon carbide layer has studied. In order to find out basic properties and characteristics, studies have conducted using various methods. Zirconium carbide is chemically vapor deposed subliming zirconium tetrachloride(ZrCl 4 ) and using methan(CH 4 ) as a source in hydrogen atmosphere. Many experiments were conducted on graphite substrate about many deposition conditions such as ZrCl 4 heating temperatures and variables of H2 and CH 4 flow rate. but carbon graphite was deposited. For deposition of zirconium carbide, several different methods were approached. so zirconium carbide deposed on ZrO 2 substrate. In this experiments. source subliming type and equipment are no problems. But deposition of zirconium carbide will be continuously studied on graphite substrate approaching views of experimental way and equipment structure

  1. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Homan, F.J.; Balthesen, E.; Turner, R.F.

    1977-01-01

    Significant advances have occurred in the development of HTGR fuel and fuel cycle. These accomplishments permit a wide choice of fuel designs, reactor concepts, and fuel cycles. Fuels capable of providing helium outlet temperatures of 750 0 C are available, and fuels capable of 1000 0 C outlet temperatures may be expected from extension of present technology. Fuels have been developed for two basic HTGR designs, one using a spherical (pebble bed) element and the other a prismatic element. Within each concept a number of variations of geometry, fuel composition, and structural materials are permitted. Potential fuel cycles include both low-enriched and high-enriched Th- 235 U, recycle Th- 233 U, and Th-Pu or U-Pu cycles. This flexibility offered by the HTGR is of great practical benefit considering the rapidly changing economics of power production. The inflation of ore prices has increased optimum conversion ratios, and increased the necessity of fuel recycle at an early date. Fuel element makeup is very similar for prismatic and spherical designs. Both use spherical fissile and fertile particles coated with combinations of pyrolytic carbon and silicon carbide. Both use carbonaceous binder materials, and graphite as the structural material. Weak-acid resin (WAR) UO 2 -UC 2 fissile fuels and sol-gel-derived ThO 2 fertile fuels have been selected for the Th- 233 U cycle in the prismatic design. Sol-gel-derived UO 2 UC 2 is the reference fissile fuel for the low-enriched pebble bed design. Both the United States and Federal Republic of Germany are developing technology for fuel cycle operations including fabrication, reprocessing, refabrication, and waste handling. Feasibility of basic processes has been established and designs developed for full-scale equipment. Fuel and fuel cycle technology provide the basis for a broad range of applications of the HTGR. Extension of the fuels to higher operating temperatures and development and commercial demonstration of fuel

  2. High Selectivity of Alkanes Production by Calcium Basic Soap Thermal Decarboxylation

    Directory of Open Access Journals (Sweden)

    Neonufa Godlief F.

    2018-01-01

    Full Text Available Renewable fuel production from vegetable oil and fat or its fatty acids by direct decarboxylation has been widely reported. An innovative approach to produce drop-in fuel via thermal catalytic decarboxylation of basic soap derived from palm stearin reported in this research. The catalytic effect of the calcium and magnesium metals in the basic soap and its decarboxylation on drop-in fuel yield and product distribution was studied. The catalytic effect was tested in the temperature range up to 370°C and atmospheric pressure for 5 hours in a batch reactor. It has been proved that the calcium basic soap decarboxylation, effectively produce the drop-in fuel in carbon ranges C8 – C20, in which more than 78% selectivity toward alkane. Whereas, only 70% selectivity toward alkane has been resulted from the magnesium basic soap decarboxylation.

  3. Surveillance of WWER-440 fuel performance

    International Nuclear Information System (INIS)

    Simko, J.; Urban, P.

    1999-01-01

    In this lecture next problems of surveillance of WWER-440 fuel performance are presented: surveillance of WWER-440 fuel performance at Mochovce NPP; basic data of WWER-440 reactor; in-core reactor measuring system 'SVRK'; basic level of SVRK; information output of basic level of SVRK; surveillance of fuel performance; table of permissible operation conditions of the reactor; limitation of the unit 1 power at the beginning of the operation; cyclic changes of power; future perspectives

  4. Notes on the next generation macromodel: a unified computations system for predicting post sealing waste depository performance

    International Nuclear Information System (INIS)

    Kaplan, S.; Kaufman, A.M.; Lathrop, J.W.

    1978-08-01

    Development and improvement activities are underway on computer software and on several analytical ideas related to the calculations of the risk from waste migration after sealing of a depository. Purpose of this report is to pull together these software and analytical ideas into an overall portrayal or blueprint of the new macromodel that is evolving, in order to clarify the relationship of the components and the interfaces required, to facilitate the alignment of all parties concerned and to allow a definitive specification of the work remaining to be done

  5. Nuclear fuel

    International Nuclear Information System (INIS)

    D Hondt, P.

    1998-01-01

    The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

  6. Development of CANFLEX fuel bundle

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan

    1991-12-01

    This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle(so-called CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactors for 1996 and 1997, and consequently will be used in the existing and future reactors in Korea. The research activities during this year include the basic design of CANFLEX fuel with slightly enriched uranium(CANFLEX-SEU), with emphasis on the extension of fuel operation limit. Based on this basic design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel. (Author)

  7. Quantifying physical characteristics of wildland fuels using the fuel characteristic classification system.

    Science.gov (United States)

    Cynthia L. Riccardi; Susan J. Prichard; David V. Sandberg; Roger D. Ottmar

    2007-01-01

    Wildland fuel characteristics are used in many applications of operational fire predictions and to understand fire effects and behaviour. Even so, there is a shortage of information on basic fuel properties and the physical characteristics of wildland fuels. The Fuel Characteristic Classification System (FCCS) builds and catalogues fuelbed descriptions based on...

  8. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    Page, R.D.

    1976-03-01

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  9. Emergency fuels utilization guidebook. Alternative Fuels Utilization Program

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    The basic concept of an emergency fuel is to safely and effectively use blends of specification fuels and hydrocarbon liquids which are free in the sense that they have been commandeered or volunteered from lower priority uses to provide critical transportation services for short-duration emergencies on the order of weeks, or perhaps months. A wide variety of liquid hydrocarbons not normally used as fuels for internal combustion engines have been categorized generically, including limited information on physical characteristics and chemical composition which might prove useful and instructive to fleet operators. Fuels covered are: gasoline and diesel fuel; alcohols; solvents; jet fuels; kerosene; heating oils; residual fuels; crude oils; vegetable oils; gaseous fuels.

  10. Fuel Cycle Concept with Advanced METMET and Composite Fuel in LWRs

    International Nuclear Information System (INIS)

    Savchenko, A.; Skupov, M.; Vatulin, A.; Glushenkov, A.; Kulakov, G.; Lipkina, K.

    2014-01-01

    The basic factor that limits the serviceability of fuel elements developing in the framework of RERTR Program (transition from HEU to LEU fuel of research reactors) is interaction between U10Mo fuel and aluminium matrix . Interaction results in extra swelling of fuels, disappearance of a heat conducting matrix, a temperature rise in the fuel centre, penetration porosity, etc. Several methods exist to prevent fuel-matrix interaction. In terms of simplifying fuel element fabrication technology and reducing interaction, doping of fuel is the most optimal version

  11. Basic concepts on threat, vulnerability and risk associated to a process of massive use of gaseous fuels

    International Nuclear Information System (INIS)

    Ospina G, J.A.

    1998-01-01

    The massive use of gaseous fuels, natural gas and liquefied gas of the oil or G.L.P., demands the implementation and complex technologies utilization in each one of the technical activities operative that transportation, distribution and consumption of these fuels requires, in order to accomplish each one of they in the form most efficient and possible insurance. Natural phenomena (earthquakes, landslides) and technical phenomena and caused by the man (terrorism, not appropriate designs of transportation and distribution net, wrong manipulation of the gas equipment) represent a latent danger for the distributors as well as for the users of the gas, generating a threat, of large proportions, for the normal development of the program of massive use of the gas, in any geographical environment where this is implemented. In this article are described the disaster concepts, threat, vulnerability and risk and the form as are related to the project of massive use of the gas, basically to transportation areas, distribution and final use

  12. High performance fuel technology development

    Energy Technology Data Exchange (ETDEWEB)

    Koon, Yang Hyun; Kim, Keon Sik; Park, Jeong Yong; Yang, Yong Sik; In, Wang Kee; Kim, Hyung Kyu [KAERI, Daejeon (Korea, Republic of)

    2012-01-15

    {omicron} Development of High Plasticity and Annular Pellet - Development of strong candidates of ultra high burn-up fuel pellets for a PCI remedy - Development of fabrication technology of annular fuel pellet {omicron} Development of High Performance Cladding Materials - Irradiation test of HANA claddings in Halden research reactor and the evaluation of the in-pile performance - Development of the final candidates for the next generation cladding materials. - Development of the manufacturing technology for the dual-cooled fuel cladding tubes. {omicron} Irradiated Fuel Performance Evaluation Technology Development - Development of performance analysis code system for the dual-cooled fuel - Development of fuel performance-proving technology {omicron} Feasibility Studies on Dual-Cooled Annular Fuel Core - Analysis on the property of a reactor core with dual-cooled fuel - Feasibility evaluation on the dual-cooled fuel core {omicron} Development of Design Technology for Dual-Cooled Fuel Structure - Definition of technical issues and invention of concept for dual-cooled fuel structure - Basic design and development of main structure components for dual- cooled fuel - Basic design of a dual-cooled fuel rod.

  13. Economic feasibility prediction of the commercial fuel cells

    International Nuclear Information System (INIS)

    Ma Yan; Karady, George G.; Winston, Anthony; Gilbert, Palomino; Hess, Robert; Pelley, Don

    2009-01-01

    This paper presents a prediction method and corresponding Visual Basic program to evaluate the economic feasibility of the commercial fuel cells in utility systems. The economic feasibility of a fuel cell is defined as having a net present value (NPV) greater than zero. The basic process of the method is to combine fuel cell specifications and real energy market data to calculate yearly earning and cost for obtaining the NPV of fuel cells. The Fuel Cell Analysis Software was developed using Visual Basic based on the proposed method. The investigation of a 250 kW molten carbonate fuel cell (FuelCell Energy DFC300A) predicted that, for application specifically in Arizona, United States, no profit would result from the installation of this fuel cell. The analysis results indicated that the efficiency, investment cost, and operation cost are three key factors affecting potential feasibility of the commercial fuel cells

  14. Seventh Edition Fuel Cell Handbook

    Energy Technology Data Exchange (ETDEWEB)

    NETL

    2004-11-01

    Provides an overview of fuel cell technology and research projects. Discusses the basic workings of fuel cells and their system components, main fuel cell types, their characteristics, and their development status, as well as a discussion of potential fuel cell applications.

  15. Features of RAPTA-SFD code modelling of chemical interactions of basic materials of the WWER active zone in accident conditions with severe fuel damage

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Sokolov, N.B.; Salatov, A.V.; Nechaeva, O.A.; Andreyeva-Andrievskaya, L.N.; Vlasov, F.Yu.

    1996-01-01

    A brief description of RAPTA-SFD code intended for computer simulations of WWER-type fuel elements (simulator or absorber element) in conditions of accident with severe damage of fuel. Presented are models of chemical interactions of basic materials of the active zone, emphasized are special feature of their application in carrying out of the CORA-W2 experiment within the framework of International Standard Problem ISP-36. Results obtained confirm expediency of phenomenological models application. (author). 6 refs, 7 figs, 1 tab

  16. Report of comment to the Nuclear Power Inspectorate concerning the final waste repository at Forsmark (SFR)

    International Nuclear Information System (INIS)

    1983-04-01

    The institute gives its support to the construction of the final depository of low and medium level radioactive waste at Forsmark. The main outline has been presented by the Swedish Nuclear Fuel Supply Company in their application. Prior to putting into operation necessary instructions have to be issued and prior to closing the depository its impact on the environment is to be examined. (G.B.)

  17. Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

    French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

  18. Report of 6th research meeting on basic process of fuel cycle for nuclear fusion reactors, Yayoi Research Group; 3rd expert committee on research of nuclear fusion fuel material correlation basis

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-01

    In this report, the lecture materials of Yayoi Research Group, 6th research meeting on basic process of fuel cycle for nuclear fusion reactors which was held at the University of Tokyo on March 25, 1996, are collected. This workshop was held also as 3rd expert committee on research of nuclear fusion fuel material correlation basis of Atomic Energy Society of Japan. This workshop has the character of the preparatory meeting for the session on `Interface effect in nuclear fusion energy system` of the international workshop `Interface effect in quantum energy system`, and 6 lectures and one comment were given. The topics were deuterium transport in Mo under deuterium ion implantation, the change of the stratum structure of graphite by hydrogen ion irradiation, the tritium behavior in opposing materials, the basic studies of the irradiation effects of solid breeding materials, the research on the behavior of hydroxyl group on the surface of solid breeding materials, the sweep gas effect on the surface of solid breeding materials, and the dynamic behavior of ion-implanted deuterium in proton-conductive oxides. (K.I.)

  19. Physical properties of granite relevant to near field conditions in a nuclear waste depository

    International Nuclear Information System (INIS)

    McLaren, J.R.; Titchell, I.

    1981-10-01

    This report presents results of the effects of heat and time at temperature on Young's modulus of granite pertinent to fission product heating in a depository for radioactive waste. In general, modulus remains constant at approximately 65 GPa to 60 to 80 0 C and then falls in nearly linear fashion to approximately 6 GPa at 550 0 C. This effect is ascribed to cracking due to differential thermal expansion between the constituent minerals or between differently oriented crystals of the same mineral. An attempt has been made to quantify the extent of cracking and hence calculate an increase in surface area exposed of between 10 2 and 10 3 m 2 .m -3 of rock by heating to 200 0 C. The effects of overburden pressure have not been studied. (author)

  20. Fuel-Coolant Interactions - some Basic Studies at the UKAEA Culham Laboratory

    International Nuclear Information System (INIS)

    Reynolds, J.A.; Dullforce, T.A.; Peckover, R.S.; Vaughan, G.J.

    1976-01-01

    In a hypothetical fault sequence important effects of fuel-coolant interactions include voiding and dispersion of core debris as well as the pressure damage usually discussed. The development of the fuel-coolant interaction probably depends on any pre-mixing Weber break-up that may occur, and is therefore a function of the way the fuel and coolant come together. Four contact modes are identified: jetting, shock tube, drops and static, and Culham's experiments have been mainly concerned with simulating the falling drop mode by using molten tin in water. It was observed that the fuel-coolant interaction is a short series of violent coolant oscillations centred at a localized position on the drop, generating a spray of submillimeter sized debris. The interaction started spontaneously at a specific time after the drop first contacted the water. There was a definite limited fuel-coolant interaction zone on a plot of initial coolant temperature versus initial fuel temperature outside which interactions never occurred. The. interaction time was a function of the initial temperatures. Theoretical scaling formulae are given which describe the fuel-coolant interaction zone and dwell time. Bounds of fuel and coolant temperature below which fuel-coolant interactions do not occur are explained by freezing. Upper bounds of fuel and coolant temperatures above which there were no fuel-coolant interactions are interpreted in terms of heat transfer through vapour films of various thicknesses. In conclusion: We have considered the effects of fuel-coolant interactions in a hypothetical fault sequence, emphasising that debris and vapour production as well as the pressure pulse can be important factors. The fuel-coolant interaction has been classified into types, according to possible modes of mixing in the fault sequence. Culham has been studying one type, the self-triggering of falling drops, by simulant experiments. It is found that there is a definite zone of interaction on a plot

  1. "Cream-skimming" in subprime mortgage securitizations : which subprime mortgage loans were sold by depository institutions prior to the crisis of 2007?

    OpenAIRE

    Paul S. Calem; Christopher Henderson; Jonathan Liles

    2010-01-01

    Depository institutions may use information advantages along dimensions not observed or considered by outside parties to "cream-skim," meaning to transfer risk to naive, uninformed, or unconcerned investors through the sale or securitization process. This paper examines whether "cream-skimming" behavior was common practice in the subprime mortgage securitization market prior to its collapse in 2007. Using Home Mortgage Disclosure Act data merged with data on subprime loan delinquency by ZIP c...

  2. Development of fuel cycles with new fuel with 8.9 mm external diameter for VVER-440. Preliminary assessment of operating efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Gagarinskiy, Alexey [National Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2017-09-15

    Since the introduction of VVERs-440, their fuel assemblies are subject to ongoing improvements. Until now, the basic structural parameters of fuel, such as rod diameter of 9.1 mm, have never changed. This paper focuses on computational estimates of basic neutronic parameters of the fuel cycle that involves assemblies consisting of fuel rods with diameter reduced to 8.9 mm.

  3. 14 CFR 31.85 - Required basic equipment.

    Science.gov (United States)

    2010-01-01

    ... following equipment is required: (a) For all balloons: (1) [Reserved] (2) An altimeter. (3) A rate of climb indicator. (b) For hot air balloons: (1) A fuel quantity gauge. If fuel cells are used, means must be... AIRWORTHINESS STANDARDS: MANNED FREE BALLOONS Operating Limitations and Information § 31.85 Required basic...

  4. Environmental protection problems in the vicinity of the Zelazny most flotation wastes depository in Poland.

    Science.gov (United States)

    Lasocki, Stanislaw; Antoniuk, Janusz; Moscicki, Jerzy

    2003-08-01

    The Zelazny Most depository of wastes from copper-ore processing, located in southwest Poland, is the largest mineral wastes repository in Europe. Moreover, it is located in a seismically active area. The seismicity is induced and is connected with mining works in the nearby underground copper mines. Any release of the contents of the repository to the environment could have devastating and even catastrophic consequences. For this reason, geophysical methods are used for continuous monitoring the state of the repository containment dams. The article presents examples of the application of geoelectric methods for detecting sites of leakage of contaminated water and a sketch of the seismic hazard analysis, which was used to predict future seismic vibrations of the repository dams.

  5. The NPOESS Preparatory Project Science Data Segment (SDS) Data Depository and Distribution Element (SD3E) System Architecture

    Science.gov (United States)

    Ho, Evelyn L.; Schweiss, Robert J.

    2008-01-01

    The National Polar-orbiting Operational Environmental Satellite System (NPOESS) Preparatory Project (NPP) Science Data Segment (SDS) will make daily data requests for approximately six terabytes of NPP science products for each of its six environmental assessment elements from the operational data providers. As a result, issues associated with duplicate data requests, data transfers of large volumes of diverse products, and data transfer failures raised concerns with respect to the network traffic and bandwidth consumption. The NPP SDS Data Depository and Distribution Element (SD3E) was developed to provide a mechanism for efficient data exchange, alleviate duplicate network traffic, and reduce operational costs.

  6. Basic experimental study with visual observation on elimination of the re-criticality issue using the MELT-II facility. Simulated fuel-escape behavior through a coolant channel

    International Nuclear Information System (INIS)

    Matsuba, Ken-ichi; Imahori, Shinji; Isozaki, Mikio

    2004-11-01

    In a core disruptive accident of fast reactors, fuel escape from the reactor core is a key phenomenon for prevention of re-criticality with significant mechanical-energy release subsequent to formation of a large-scale fuel pool with high mobility. Therefore, it is effective to study possibility of early fuel escape through probable escape paths such as a control-rod-guide-tube space well before high-mobility-pool formation. The purpose of the present basic experimental study is to clarify the mechanism of fuel-escape under a condition expected in the reactor situation, in which some amount of coolant may be entrapped into the molten-fuel pool. The following results have been obtained through basic experiments in which molten Wood's metal (components: 60wt%Bi-20wt%Sn-20wt%In, density at the room temperature: 8700 kg/m 3 , melting point: 78.8degC) is ejected into an coolant channel filled with water. (1) In the course of melt ejection, a small quantity of coolant is forced to be entrapped into the melt pool as a result of thermal interactions leading to high-pressure rise within the coolant channel. (2) Melt ejection is accelerated by pressure build-up which results from vapor pressure of entrapped coolant within the melt pool. (3) Average melt-ejection rate tends to increase in lower coolant-subcooling conditions, in which pressure build-up within the melt pool is enhanced. These results indicate a probability of a phenomenon in which melt ejection is accelerated by entrapment of coolant within a melt pool. Through application of the mechanism of confirmed phenomenon into the reactor condition, it is suggested that fuel escape is enhanced by entrapment of coolant within a fuel pool. (author)

  7. Basic properties of fuel determining its behavior under irradiation

    International Nuclear Information System (INIS)

    Konovalov, I.I.

    2000-01-01

    The theoretical model describing a swelling of nuclear fuel at low irradiation temperatures is considered. The critical physical parameters of substances determining behavior of point defects, gas fission atoms, dislocation density, nucleation and growth of gas-contained pores are determined. The correlation between meanings of critical parameters and physical properties of substance is offered. The accounts of swelling of various dense fuels with reference to work in conditions of research reactors are given. (author)

  8. Safety research basic plan of JNC

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-01

    Japan Nuclear Cycle Development Institute (JNC) formally succeeded to Power Reactor and Nuclear Fuel Development Corporation (PNC) on October, 1 1998. This report describes the basic plan for major program of JNC which consists of two parts: management philosophy of the new institute and the latest revised medium term program. In the first part, the primary mission of JNC is to perform its R and D concentrating on fast breeder reactor and its fuel cycle, and treatment and disposal of high-level radioactive wastes, while at the same time giving special consideration to safety. In the second, individual programs in the new basic plan are discussed in detail. The outline and schedule of each program are also attached in the table form. (H. Itami)

  9. Nuclear fuel storage

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1981-01-01

    A nuclear fuel storage apparatus for use in a water-filled pool is fabricated of a material such as stainless steel in the form of an egg crate structure having vertically extending openings. Fuel may be stored in this basic structure in a checkerboard pattern with high enrichment fuel, or in all openings when the fuel is of low effective enrichment. Inserts of a material such as stainless steel are adapted to fit within these openings so that a water gap and, therefore, a flux trap is formed between adjacent fuel storage locations. These inserts may be added at a later time and fuel of a higher enrichment may be stored in each opening. When it is desired to store fuel of still greater enrichment, poison plates may be added to the water gap formed by the installed insert plates, or substituted for the insert plates. Alternately, or in addition, fuel may be installed in high neutron absorption poison boxes which surround the fuel assembly. The stainless steel inserts and the poison plates are each not required until the capacity of the basic egg crate structure is approached. Purchase of these items can, therefore, be deferred for many years. Should the fuel to be stored be of higher enrichment than initially forecast, the deferred decision on the poison plates makes it possible to obtain increased poison in the plates to satisfy the newly discovered requirement

  10. Automotive Fuel and Exhaust Systems.

    Science.gov (United States)

    Irby, James F.; And Others

    Materials are provided for a 14-hour course designed to introduce the automotive mechanic to the basic operations of automotive fuel and exhaust systems incorporated on military vehicles. The four study units cover characteristics of fuels, gasoline fuel system, diesel fuel systems, and exhaust system. Each study unit begins with a general…

  11. BASIC program to compute uranium density and void volume fraction in laboratory-scale uranium silicide aluminum dispersion plate-type fuel

    International Nuclear Information System (INIS)

    Ugajin, Mitsuhiro

    1991-05-01

    BASIC program simple and easy to operate has been developed to compute uranium density and void volume fraction for laboratory-scale uranium silicide aluminum dispersion plate-type fuel, so called miniplate. An example of the result of calculation is given in order to demonstrate how the calculated void fraction correlates with the microstructural distribution of the void in a miniplate prepared in our laboratory. The program is also able to constitute data base on important parameters for miniplates from experimentally-determined values of density, weight of each constituent and dimensions of miniplates. Utility programs pertinent to the development of the BASIC program are also given which run in the popular MS-DOS environment. All the source lists are attached and brief description for each program is made. (author)

  12. Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P.O. 1236909 Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

    1978-12-01

    The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel

  13. Advanced Technology and Alternative Fuel Vehicles

    International Nuclear Information System (INIS)

    Tuttle, J.

    2001-01-01

    This fact sheet provides a basic overview of today's alternative fuel choices--including biofuels, biodiesel, electricity, and hydrogen--alternative fuel vehicles, and advanced vehicle technology, such as hybrid electric vehicles, fuel cells and advanced drive trains

  14. Basic research on high-uranium density fuels for research and test reactors

    International Nuclear Information System (INIS)

    Ugajin, M.; Itoh, A.; Akabori, M.

    1992-01-01

    High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)

  15. 1989 basic plan for atomic energy development and utilization

    International Nuclear Information System (INIS)

    1989-01-01

    A Basic Plan for Atomic Energy Development and Utilization has been established each year based on the guidelines set up by the Atomic Energy Commission of Japan, with the aim of promoting the development and utilization of atomic energy schematically and efficiently. The Basic Plan shows specific projects to achieve the objectives specified in the Long-Range Plan for Atomic Energy Development and Utilization. The Basic Plan specifies efforts to be made for overall strengthening of safety measures (safety policies, safety research, disaster prevention, etc.), promotion of nuclear power generation, establishment of the nuclear fuel cycle (securing of uranium, technology for uranium enrichment, reprocessing, etc.), development of new types of power reactors (fast breeder reactor, new types of converter reactors, plutonium fuel processing technology), promotion of leading projects (nuclear fusion, utilization of radiations, atomic powered ships, high-temperature engineering tests), promotion of basic technology development (basic research, training of scientists and engineers), voluntary and active international activities (international cooperation), and acquisition of understanding and cooperation of the general public. (N,K.)

  16. Development of generalized boiling transition model applicable for wide variety of fuel bundle geometries. Basic strategy and numerical approaches

    International Nuclear Information System (INIS)

    Ninokata, Hisashi; Sadatomi, Michio; Okawa, Tomio

    2003-01-01

    In order to establish a key technology to realize advanced BWR fuel designs, a three-year project of the advanced subchannel analysis code development had been started since 2002. The five dominant factors involved in the boiling transitional process in the fuel bundles were focused. They are, (1) inter-subchannel exchanges, (2) influences of obstacles (3) dryout of liquid film, (4) transition of two-phase flow regimes and (5) deposition of droplets. It has been recognized that present physical models or constitutive equations in subchannel formulations need to be improved so that they include geometrical effects in the fuel bundle design more mechanistically and universally. Through reviewing literatures and existent experimental results, underlying elementary processes and geometrical factors that are indispensable for improving subchannel codes were identified. The basic strategy that combines numerical and experimental approaches was proposed aiming at establishment of mechanistic models for the five dominant factors. In this paper, the present status of methodologies for detailed two-phase flow studies has been summarized. According to spatial scales of focused elementary processes, proper numerical approaches were selected. For some promising numerical approaches, preliminary calcitonins were performed for assessing their applicability to investigation of elementary processes involved in the boiling transition. (author)

  17. Uranium: a basic evaluation

    International Nuclear Information System (INIS)

    Crull, A.W.

    1978-01-01

    All energy sources and technologies, including uranium and the nuclear industry, are needed to provide power. Public misunderstanding of the nature of uranium and how it works as a fuel may jeopardize nuclear energy as a major option. Basic chemical facts about uranium ore and uranium fuel technology are presented. Some of the major policy decisions that must be made include the enrichment, stockpiling, and pricing of uranium. Investigations and lawsuits pertaining to uranium markets are reviewed, and the point is made that oil companies will probably have to divest their non-oil energy activities. Recommendations for nuclear policies that have been made by the General Accounting Office are discussed briefly

  18. Development of design technology for dual-cooled fuel

    International Nuclear Information System (INIS)

    Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Young Ho

    2010-03-01

    Primary purpose of the project is to complete a basic design of the power uprating dual-cooled fuel's structural components for an actual use in the existing nuclear power plants. It also includes a basic design of the components of a dual-cooled fuel rod. To this end, during the three years of the first stage (2007.03.∼2010.02.), concepts and technical issues of the structural components such as a supporting structure, guide thimbles and instrumentation tube and the top and bottom end pieces were derived in order to comply with the functional requirements and design criteria of them. Basic design was carried out to resolve the issues by using analytical methods as well as experiments, and observed finally is that a structural compatibility of the designed dual-cooled fuel to the Korean Standard Nuclear Power Plant (OPR-1000). As for the dual-cooled fuel rod's components such as a plenum spring, a spacer and end plugs, a concept of them was established by using the basic dimension and array produced by other sub-projects. In turn, the basic design was completed by using the finite element analysis and conventional mechanical design formulae. Additionally, a welding method and equipment for a dual-cooled fuel rod specimen was also successfully developed to prepare for the irradiation tests at the HANARO. It was shown that a dual-cooed fuel for the OPR-1000 can be designed after manufacturing the partial assembly with the designed components and their drawings. The first stage was completed with passing the Gate checks proposed at the beginning. During the second stage(2010.03.∼2012.02.), researches on the mechanical behavior and structural integrity of the designed dual-cooled fuel will be conducted for preparing a license of it, which should be done when the dual-cooled fuel is commercialized

  19. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

    1980-01-01

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  20. Conceptual design of retrieval systems for emplaced transuranic waste containers in a salt bed depository. Final report

    International Nuclear Information System (INIS)

    Fogleman, S.F.

    1980-04-01

    The US Department of Energy and the Nuclear Regulatory Commission have jurisdiction over the nuclear waste management program. Design studies were previously made of proposed repository site configurations for the receiving, processing, and storage of nuclear wastes. However, these studies did not provide operational designs that were suitable for highly reliable TRU retrieval in the deep geologic salt environment for the required 60-year period. The purpose of this report is to develop a conceptual design of a baseline retrieval system for emplaced transuranic waste containers in a salt bed depository. The conceptual design is to serve as a working model for the analysis of the performance available from the current state-of-the-art equipment and systems. Suggested regulations would be based upon the results of the performance analyses

  1. Extending the basic function of lattice oxygen in lepidocrocite titanate - The conversion of intercalated fatty acid to liquid hydrocarbon fuels

    Science.gov (United States)

    Maluangnont, Tosapol; Arsa, Pornanan; Sooknoi, Tawan

    2017-12-01

    We report herein the basicity of the external and internal lattice oxygen (OL) in lepidocrocite titanates with respect to CO2 and palmitic acid, respectively. Several compositions have been tested with different types of the metal M aliovalently (co)substituted for Ti, K0.8[MyTi2-y]O4 (M = Li, Mg, Fe, Co, Ni, Cu, Zn, Cu/Ni and Cu/Zn). The low CO2 desorption peak temperature (70-100 °C) suggests that the external OL sites are weakly basic similar to TiO2. However, the internal OL sites are sufficiently basic to deprotonate palmitic acid, forming the intercalated potassium palmitate at the interlayer spaces. The latter serves as a two-dimensional (2D) molecular reactor for the production of liquid hydrocarbon fuels via deoxygenation under atmospheric N2. A relationship has been observed between the yield of the liquid products vs the partial charge of the lattice oxygen (δO). Since the deoxygenation pathway is highly dependent on the metal substitution, the redox-active sites might also play some roles. The co-substituted K0.8[Cu0.2Ni0.2]Ti1.6O4 produced 68.0% yield of the liquid products, with 51% saturated and 15% unsaturated C15 hydrocarbons at 350 °C.

  2. Systematic study on Thorium fuel

    International Nuclear Information System (INIS)

    Shibata, Toshikazu; Kimura, Itsuro; Iwata, Shiro; Furuya, Hirotaka; Suzuki, Susumu.

    1988-01-01

    Introduced is the activities of the Joint Research Project Team on Thorium Fuel organized by mainly university researchers in Japan and supported by the Ministry of Education, Science and Culture for seven years since 1980. Four major groups were organized; (1) nuclear data, reactor physics and design, (2) nuclear fuel, (3) down stream and (4) biological effects of thorium. The first group covered measurements and analysis on nuclear data of thorium related nuclides, experiment and analysis on nuclear characteristics of thorium containing cores, basic engineering on a thorium molten salt reactor, and designs of several types of reactors. Fabrication and irradiation tests of thorium oxide fuel, and basic studies on new type thorium fuels (e.g. carbide and nitride) were studied by the second group. The third group covered the use of solutions in reprocessing of spent fuel, behavior of fission products, immobilization of high level radioactive waste, and continuous reprocessing for a molten salt reactor. The fourth group performed the trace study for patients who had been intravascularly injected with thorotrast for diagnosis of war injuries during the Second World War. (author)

  3. The basic concepts of a fuel-power detector for nuclear power reactors

    International Nuclear Information System (INIS)

    Lynch, G.F.

    1979-01-01

    Fuel power is proposed as an alternative to neutron or gamma-ray flux for control and safety functions in CANDU power reactors. To satisfy in-core power monitoring requirements, a detector whose dynamic response corresponds to the heat production rate in the fuel is needed. This report explores the concept of tailoring the response characteristics of a mixed-response self-powered flux detector to match the requirements of an ideal fuel-power detector. (author)

  4. Photoactivated Fuel Cells (PhotoFuelCells. An alternative source of renewable energy with environmental benefits

    Directory of Open Access Journals (Sweden)

    Stavroula Sfaelou

    2016-03-01

    Full Text Available This work is a short review of Photoactivated Fuel Cells, that is, photoelectrochemical cells which consume an organic or inorganic fuel to produce renewable electricity or hydrogen. The work presents the basic features of photoactivated fuel cells, their modes of operation, the materials, which are frequently used for their construction and some ideas of cell design both for electricity and solar hydrogen production. Water splitting is treated as a special case of photoactivated fuel cell operation.

  5. Ethanol Fuels Reference Guide: A Decision-Makers Guide to Ethanol Fuels

    Energy Technology Data Exchange (ETDEWEB)

    1982-10-01

    This guide is a compendium of information on alcohol fuel production and use. Chapter titles are: facts about ethanol; gasohol-answers to the basic questions; feedstocks and their coproducts; ethanol production processes; and vehicle fuel use and performance. In addition, there are 8 appendices which include fermentation guides for common grains and potatoes, component and enzyme manufacturers, and information on regulations and permits. (DMC)

  6. Study on the performance of fuel elements with carbide and carbide-nitride fuel

    International Nuclear Information System (INIS)

    Golovchenko, Yu.M.; Davydov, E.F.; Maershin, A.A.

    1985-01-01

    Characteristics, test conditions and basic results of material testing of fuel elements with carbide and carbonitride fuel irradiated in the BOR-60 reactor up to 3-10% burn-up at specific power rate of 55-70 kW/m and temperatures of the cladding up to 720 deg C are described. Increase of cladding diameter is stated mainly to result from pressure of swelling fuel. The influence of initial efficient porosity of the fuel on cladding deformation and fuel stoichiometry on steel carbonization is considered. Utilization of carbide and carbonitride fuel at efficient porosity of 20% at the given test modes is shown to ensure their operability up to 10% burn-up

  7. Structure of fuel performance audit code for SFR metal fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong Sik; Kim, Hyo Chan [KAERI, Daejeon (Korea, Republic of); Jeong, Hye Dong; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-10-15

    A Sodium Cooled Fast Reactor (SFR) is a promising option to solve the spent fuel problems, but, there are still much technical issues to commercialize a SFR. One of issues is a development of advanced fuel which can solve the safety and the economic issues at the same time. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured. In Korea Institute of Nuclear Safety (KINS), the new project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. To develop the new code system, the code structure design and its requirements need to be studied. Various performance models and code systems are reviewed and their characteristics are analyzed in this paper. Based on this study, the fundamental performance models are deduced and basic code requirements and structure are established.

  8. Recent situation of the establishment of nuclear fuel cycle

    International Nuclear Information System (INIS)

    Hoshiba, Shizuo

    1982-01-01

    In Japan, the development of nuclear power as principal petroleum substitute is actively pursued. Nuclear power generation now accounts for about 17 % of the total power generation in Japan. The business related to nuclear fuel cycle should be established by private enterprises. The basic policy in the establishment of nuclear fuel cycle is the stabilized supply of natural uranium, raise in domestic production of enriched uranium, dFomestic fuel reprocessing in principle, positive plutonium utilization, and so on. After explaining this basic policy, the present situation and problems in the establishment of nuclear fuel cycle are described: securing of uranium resources, securing of enriched uranium, reprocessing of used fuel, utilization of plutonium, management of radioactive wastes. (Mori, K.)

  9. Basic planning of a newly built exclusive ship for spent fuel transport

    International Nuclear Information System (INIS)

    Obara, I.; Sasao, T.; Akiyama, H.; Kybota, T.

    1998-01-01

    A commercial reprocessing plant is under construction at the Fuel Cycle Facilities in Rokkasho-mura, Aomori Prefecture. To prepare for the transport of spent nuclear fuels (SF) from all Japanese nuclear power stations to this reprocessing plant, the need for an exclusive transport ship was recognized. Nuclear Fuel Transport Co. Ltd. (NFT), in cooperation with electric power utilities planned the construction of such a ship over a period of several years. During this period NFT developed new types of cask to transport high burn-up spent fuels to the reprocessing plant. Six kinds of casks were developed and 40 units are now under fabrication. The ship was designed to carry a maximum of 20 units. Based on the Irradiated Nuclear Fuel (INF) Code adopted by the International Maritime Organization (IMO), the Japanese Ministry of Transport (MoT) issued new domestic regulations in September, 1995 which covered design criteria for ships carrying Irradiated Nuclear Fuels. The new SF transport ship is the first one to which this new regulation was applied. Although the ship will only ply the coastal routes of Japan, she has been designed to conform with all the international requirements for the Class-3 of the INF Code. In May 1995, Nuclear Fuel Shipping Co. Ltd (NFS), a wholly-owned subsidiary of NFT, concluded a contract with Mitsui Engineering and Shipbuilding Co., Ltd. for the construction of the exclusive transport ship. The keel was laid in November 1995. The ship was launched in april 1996 and named 'Rokuei-Maru'. At the end of September, she was completed and delivered to the ship owner, NFS. (authors)

  10. Joint convention on the safety of spent fuel management and on the safety of radioactive waste management. First national report on the implementation by France of the obligations of the Convention

    International Nuclear Information System (INIS)

    2003-03-01

    The Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management is supplementing the Convention of Nuclear Safety. it was approved by France on february 22, 2000 and it entered into force on June 18,2001. Article 32 obliges each contracting Party to present at the review meetings (every three years) a report on the way in which it implements the obligations of the Convention (full text of the Convention and additional information on the web site of the IAEA, its director General being the depository of the Convention. (author)

  11. Poor access to basic services | IDRC - International Development ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    2015-10-28

    Oct 28, 2015 ... Poor access to basic services can foster competition and fuel conflict between groups. ... Make clean water, sanitation, electricity, and other services accessible ... Poverty, inequality, and violence in urban India: Towards more ...

  12. Conversion of bio-feedstocks through acid and basic zeolites and catalysts

    Energy Technology Data Exchange (ETDEWEB)

    Buzzoni, R.; Bosetti, A.; Delledonne, D.; Perego, C. [eni S.p.A. Research Centre for Non-Conventional Energy, Novara (Italy). Ist. eni Donegani

    2012-07-01

    Not far in the future, a significant part of fuels and chemicals will be originated by renewable biomass resources. In this respect, zeolite catalysts may help to develop a new generation of bio-fuel and chemical processes. In the new bio-paradigm not only acid but also basic materials will have an important and dominant role. Just to give some examples, basic zeolites based catalysts have been proposed for transesterification of triglyceride esters of fatty acids to biodiesel, for disrupting the lignin polymer by base catalyzed depolymerisation and for one pot lignin liquefaction by hydrogenation. (orig.)

  13. Mixed U/Pu oxide fuel fabrication facility co-processed feed, pelletized fuel

    International Nuclear Information System (INIS)

    1978-09-01

    Two conceptual MOX fuel fabrication facilities are discussed in this study. The first facility in the main body of the report is for the fabrication of LWR uranium dioxide - plutonium dioxide (MOX) fuel using co-processed feed. The second facility in the addendum is for the fabrication of co-processed MOX fuel spiked with 60 Co. Both facilities produce pellet fuel. The spiked facility uses the same basic fabrication process as the conventional MOX plant but the fuel feed incorporates a high energy gamma emitter as a safeguard measure against diversion; additional shielding is added to protect personnel from radiation exposure, all operations are automated and remote, and normal maintenance is performed remotely. The report describes the fuel fabrication process and plant layout including scrap and waste processing; and maintenance, ventilation and safety measures

  14. International guidelines for fire protection at nuclear installations including nuclear fuel plants, nuclear fuel stores, teaching reactors, research establishments

    International Nuclear Information System (INIS)

    The guidelines are recommended to designers, constructors, operators and insurers of nuclear fuel plants and other facilities using significant quantities of radioactive materials including research and teaching reactor installations where the reactors generally operate at less than approximately 10 MW(th). Recommendations for elementary precautions against fire risk at nuclear installations are followed by appendices on more specific topics. These cover: fire protection management and organization; precautions against loss during construction alterations and maintenance; basic fire protection for nuclear fuel plants; storage and nuclear fuel; and basic fire protection for research and training establishments. There are numerous illustrations of facilities referred to in the text. (U.K.)

  15. Thorium-based nuclear fuel: current status and perspectives

    International Nuclear Information System (INIS)

    1987-03-01

    Until the present time considerable efforts have already been made in the area of fabrication, utilization and reprocessing of Th-based fuels for different types of reactors, namely: by FRG and USA - for HTRs; FRG and Brazil, Italy - for LWRs; India - for HWRs and FBRs. Basic research of thorium fuels and thorium fuel cycles are also being undertaken by Australia, Canada, China, France, FRG, Romania, USSR and other countries. Main emphasis has been given to the utilization of thorium fuels in once-through nuclear fuel cycles, but in some projects closed thorium-uranium or thorium-plutonium fuel cycles are also considered. The purpose of the Technical Committee on the Utilization of Thorium-Based Nuclear Fuel: Current Status and Perspective was to review the world thorium resources, incentives for further exploration, obtained experience in the utilization of Th-based fuels in different types of reactors, basic research, fabrication and reprocessing of Th-based fuels. As a result of the panel discussion the recommendations on future Agency activities and list of major worldwide activities in the area of Th-based fuel were developed. A separate abstract was prepared for each of the 9 papers in this proceedings series

  16. Preliminary Nuclear Analysis for the HANARO Fuel Element with Burnable Absorber

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, So Young; In, Won Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Burnable absorber is used for reducing reactivity swing and power peaking in high performance research reactors. Development of the HANARO fuel element with burnable absorber was started in the U-Mo fuel development program at HANARO, but detailed full core analysis was not performed because the current HANARO fuel management system is uncertain to analysis the HANARO core with burnable absorber. A sophisticated reactor physics system is required to analysis the core. The McCARD code was selected and the detailed McCARD core models, in which the basic HANARO core model was developed by one of the McCARD developers, are used in this study. The development of nuclear fuel requires a long time and correct developing direction especially by the nuclear analysis. This paper presents a preliminary nuclear analysis to promote the fuel development. Based on the developed fuel, the further nuclear analysis will improve reactor performance and safety. Basic nuclear analysis for the HANARO and the AHR were performed for getting the proper fuel elements with burnable absorber. Addition of 0.3 - 0.4% Cd to the fuel meat is promising for the current HANARO fuel element. Small addition of burnable absorber may not change any fuel characteristics of the HANARO fuel element, but various basic tests and irradiation tests at the HANARO core are required.

  17. Enhanced wood fuel handling: market and design studies

    Energy Technology Data Exchange (ETDEWEB)

    Landen, R.; Rippengal, R.; Redman, A.N.

    1997-09-01

    This report examines the potential for the manufacture and sale of novel wood fuel handling systems as a means of addressing users' concerns regarding current capital costs and potential high labour costs of non-automated systems. The report considers fuel handling technology that is basically appropriate for wood-fired heating systems of between c.100kW and c.1MW maximum continuous rating. This report details work done by the project collaborators in order to: (1) assess the current status of wood fuel handling technology; (2) evaluate the market appetite for improved wood fuel handling technology; (3) derive capital costs which are acceptable to customers; (4) review design options; and (5) select one or more design options worthy of further development. The current status of wood fuel handling technology is determined, and some basic modelling to give guidance on acceptable capital costs of 100-1000kW wood fuel handling systems is undertaken. (author)

  18. The next generation fuel cells: anion exchange membrane fuel cells (AEMFC)

    International Nuclear Information System (INIS)

    Tauqir, A.; Zahoor, S.

    2013-01-01

    Many environmentally friendly alternatives (solar, wind, hydroelectric, and geothermal power) can only be used in particular environments. In contrast, fuel cells can have near-zero emissions, are quiet and efficient, and can work in any environment where the temperature is lower than the cell's operating temperature. Among various types of fuel cells, the AEMFC is the most recent one and has advantages such as excellent performance compared to other candidate fuel cells due to its active O/sub 2/ electrode kinetics and flexibility to use a wide range of electro-catalysts such as silver and nickels contrary to expensive one (Platinum) required for proton exchange membrane fuel cell (PEMFC). Anion exchange membrane (AEM) is a crucial part in AEMFC, determining durability and electrochemical performances of membrane electrode assembly (MEA). The role of an AEM is to conduct hydroxyl ions from cathode to anode. If this conduction is not sufficiently high and selective, the corresponding fuel cell will not find any practical application. One of the major problems associated with AEMFC is much lower conductivities of anion compare to proton conductivity in PEMFCs, even upon similar working condition. Thus AEMs is only practical, if it is chemically and mechanically stable against severe basic operation conditions and highly hydroxyl ions conductive. The conventional AEMs based on animated aliphatic and aromatic hydrocarbon or even fluorinated polymers tend to be attacked by hydroxyl ions, causing the degradation during operation is strongly basic conditions. (author)

  19. Novel materials for fuel cells operating on liquid fuels

    Directory of Open Access Journals (Sweden)

    César A. C. Sequeira

    2017-05-01

    Full Text Available Towards commercialization of fuel cell products in the coming years, the fuel cell systems are being redefined by means of lowering costs of basic elements, such as electrolytes and membranes, electrode and catalyst materials, as well as of increasing power density and long-term stability. Among different kinds of fuel cells, low-temperature polymer electrolyte membrane fuel cells (PEMFCs are of major importance, but their problems related to hydrogen storage and distribution are forcing the development of liquid fuels such as methanol, ethanol, sodium borohydride and ammonia. In respect to hydrogen, methanol is cheaper, easier to handle, transport and store, and has a high theoretical energy density. The second most studied liquid fuel is ethanol, but it is necessary to note that the highest theoretically energy conversion efficiency should be reached in a cell operating on sodium borohydride alkaline solution. It is clear that proper solutions need to be developed, by using novel catalysts, namely nanostructured single phase and composite materials, oxidant enrichment technologies and catalytic activity increasing. In this paper these main directions will be considered.

  20. Design of fuel loading for Bohunice V-1 Unit 2 reaktor for fuel cycle No.19

    International Nuclear Information System (INIS)

    Majercik, J.

    1998-01-01

    The report contains description of the design of fuel loading for the fuel cycle No. 19 in the V-1 Bohunice Unit 2 reactor. Input data and computer codes used for the development of the design are shown. The fuel loading is characterized by the assortment of the fuel loaded and by the scheme of re shuffling of assemblies in the core. An evaluation of basic neutronic core parameters as relates to the compliance with safety criteria is a part of the report as well

  1. Operation results of 3-rd generation nuclear fuel WWER-440 in initial period

    International Nuclear Information System (INIS)

    Adeev, V.; Panov, A.

    2011-01-01

    On unit 4 of Kola NPP trial operation of 3-rd generation's fuel began in 2010. Fuel assemblies of 3-rd generation (FA-3) have a number of design features that provide better operational characteristics. Concise description of a design and the basic advantages of fuel of 3-rd generation are described in articles. Increasing of efficiency of nuclear fuel usage will be achieved by reduction of the parasitic capture of thermal neutrons in constructional materials (weight of zirconium is reduced), optimization of uranium-water relation (increase in fuel elements step), increasing of uranium loading (usage of fuel pellets with increased diameter and without central hole in them). By results of trial operation mass transition to use of given type of assemblies in WWER-440 is possible. This report presents the basic outcomes of the trial operation, a brief survey of the obtained data. The basic characteristics of the reactor core with fuel of 3-rd generation are resulted in work. (authors)

  2. Nuclear Fuel Cycle Objectives

    International Nuclear Information System (INIS)

    2013-01-01

    . The four Objectives publications include Nuclear General Objectives, Nuclear Power Objectives, Nuclear Fuel Cycle Objectives, and Radioactive Waste management and Decommissioning Objectives. This publication sets out the objectives that need to be achieved in the area of the nuclear fuel cycle to ensure that the Nuclear Energy Basic Principles are satisfied. Within each of these four Objectives publications, the individual topics that make up each area are addressed. The five topics included in this publication are: resources; fuel engineering and performance; spent fuel management and reprocessing; fuel cycles; and the research reactor nuclear fuel cycle

  3. TopFuel 2003 conference report

    International Nuclear Information System (INIS)

    Anon.

    2003-01-01

    The international conference, TopFuel 2003 - Nuclear Fuel for Today and Tomorrow, Experience and Outlook, was held in Wuerzburg on March 16-19, 2003. The event, which was organized jointly by the Atomic Energy Society of Japan (AESJ), the American Nuclear Society (ANS), the German Nuclear Society and the European Nuclear Society (ENS), provided a comprehensive overview of current topics and developments in nuclear fuel supply in more than ninety papers and poster presentations. At the plenary session, more than 300 participants from 15 countries discussed basic problems of nuclear fuel development, safety research, strategies of nuclear fuel supply in the 21st century, fuel fabrication, interim storage of fuel elements, and problems of fuel element design for nuclear power plants of the next generation. Seven technical sessions dealt with other topical developments in these fields: - feedback of experience in fuel use, - nuclear fuel cycle efforts to increase burnup, - trends in nuclear fuel design, - advanced methods and codes, - fabrication, - transport, nuclear fuel services. (orig.) [de

  4. Basic nuclear data for FBR fuel cycle. Balance and forecasting

    International Nuclear Information System (INIS)

    Costa, L.; Granget, G.; Josso, F.

    1982-01-01

    A balance is made of nuclear data needed for studying FBR fuel cycle. From the accuracy of the obtained data, sensitivity calculations have enabled the future experimental measurements to be established [fr

  5. Fast breeder fuel cycle

    International Nuclear Information System (INIS)

    1978-09-01

    Basic elements of the ex-reactor part of the fuel cycle (reprocessing, fabrication, waste handling and transportation) are described. Possible technical and proliferation measures are evaluated, including current methods of accountability, surveillance and protection. The reference oxide based cycle and advanced cycles based on carbide and metallic fuels are considered utilizing conventional processes; advanced nonaqueous reprocessing is also considered. This contribution provides a comprehensive data base for evaluation of proliferation risks

  6. The failure mechanisms of HTR coated particle fuel and computer code

    International Nuclear Information System (INIS)

    Yang Lin; Liu Bing; Shao Youlin; Liang Tongxiang; Tang Chunhe

    2010-01-01

    The basic constituent unit of fuel element in HTR is ceramic coated particle fuel. And the performance of coated particle fuel determines the safety of HTR. In addition to the traditional detection of radiation experiments, establishing computer code is of great significance to the research. This paper mainly introduces the structure and the failure mechanism of TRISO-coated particle fuel, as well as a few basic assumptions,principles and characteristics of some existed main overseas codes. Meanwhile, this paper has proposed direction of future research by comparing the advantages and disadvantages of several computer codes. (authors)

  7. Nuclear fuel cycle information workshop

    International Nuclear Information System (INIS)

    1983-01-01

    This overview of the nuclear fuel cycle is divided into three parts. First, is a brief discussion of the basic principles of how nuclear reactors work; second, is a look at the major types of nuclear reactors being used and world-wide nuclear capacity; and third, is an overview of the nuclear fuel cycle and the present industrial capability in the US

  8. Back end of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Shapar, H.K.

    1986-01-01

    Most of the nuclear spent fuel that is discharged from the reactors in OECD countries is destined currently for long term interim storage before final processing or direct disposal. There are at least three basic considerations affecting the dicision on spent fuel, that is, the capacity of prompt reprocessing is insufficient at present, reprocessing is not urgent for the reason of economy or plutonium availability, and the cooling of spent fuel in controlled storage is economically advantageous. The basic technology of reprocessing has been commercially available for several decades, but political problems and the lack of immediate incentive for reprocessing slowed the buildup of new capacity. To avoid the problems related to plutonium storage, it is reasonable to postpone reprocessing. Some OECD countries plan the direct disposal of spent fuel elements instead of reprocessing. The technology, supply and demand and cost of the storage and transport of spent fuel, reprocessing and waste disposal are discussed. The share of the back end in the total levelized fuel cycle cost is expected to be between 10 and 20 %. The impact of the choice of back end options on the cost of power generation will be only 2 %. (Kako, I.)

  9. Thermodynamic analysis of biofuels as fuels for high temperature fuel cells

    Science.gov (United States)

    Milewski, Jarosław; Bujalski, Wojciech; Lewandowski, Janusz

    2013-02-01

    Based on mathematical modeling and numerical simulations, applicativity of various biofuels on high temperature fuel cell performance are presented. Governing equations of high temperature fuel cell modeling are given. Adequate simulators of both solid oxide fuel cell (SOFC) and molten carbonate fuel cell (MCFC) have been done and described. Performance of these fuel cells with different biofuels is shown. Some characteristics are given and described. Advantages and disadvantages of various biofuels from the system performance point of view are pointed out. An analysis of various biofuels as potential fuels for SOFC and MCFC is presented. The results are compared with both methane and hydrogen as the reference fuels. The biofuels are characterized by both lower efficiency and lower fuel utilization factors compared with methane. The presented results are based on a 0D mathematical model in the design point calculation. The governing equations of the model are also presented. Technical and financial analysis of high temperature fuel cells (SOFC and MCFC) are shown. High temperature fuel cells can be fed by biofuels like: biogas, bioethanol, and biomethanol. Operational costs and possible incomes of those installation types were estimated and analyzed. A comparison against classic power generation units is shown. A basic indicator net present value (NPV) for projects was estimated and commented.

  10. Thermodynamic analysis of biofuels as fuels for high temperature fuel cells

    Directory of Open Access Journals (Sweden)

    Milewski Jarosław

    2013-02-01

    Full Text Available Based on mathematical modeling and numerical simulations, applicativity of various biofuels on high temperature fuel cell performance are presented. Governing equations of high temperature fuel cell modeling are given. Adequate simulators of both solid oxide fuel cell (SOFC and molten carbonate fuel cell (MCFC have been done and described. Performance of these fuel cells with different biofuels is shown. Some characteristics are given and described. Advantages and disadvantages of various biofuels from the system performance point of view are pointed out. An analysis of various biofuels as potential fuels for SOFC and MCFC is presented. The results are compared with both methane and hydrogen as the reference fuels. The biofuels are characterized by both lower efficiency and lower fuel utilization factors compared with methane. The presented results are based on a 0D mathematical model in the design point calculation. The governing equations of the model are also presented. Technical and financial analysis of high temperature fuel cells (SOFC and MCFC are shown. High temperature fuel cells can be fed by biofuels like: biogas, bioethanol, and biomethanol. Operational costs and possible incomes of those installation types were estimated and analyzed. A comparison against classic power generation units is shown. A basic indicator net present value (NPV for projects was estimated and commented.

  11. Nuclear design of APSARA reload-2 fuel

    International Nuclear Information System (INIS)

    Nath, M.; Veeraraghavan, N.

    1978-01-01

    In view of the satisfactory operating performance of initial and reload-1 fuel designs of Apsara reactor, it was felt desirable to adopt a basically similar design for reload-2 fuel, i.e. the fuel assembly should consist of equally spaced parallel fuel plates in which highly enriched uranium, alloyed with aluminium, is employed as fuel. However, because of fabricational constraints, certain modifications were necessary and were incorporated in the proposed reload design to cater to the multiple needs of operational requirements, improved fuel utilization and inherent reactor safety. The salient features of the nuclear design of reload-2 fuel for the Apsara reactor are discussed. (author)

  12. Cost reductions of fuel cells for transport applications: fuel processing options

    Energy Technology Data Exchange (ETDEWEB)

    Teagan, W P; Bentley, J; Barnett, B [Arthur D. Little, Inc., Cambridge, MA (United States)

    1998-03-15

    The highly favorable efficiency/environmental characteristics of fuel cell technologies have now been verified by virtue of recent and ongoing field experience. The key issue regarding the timing and extent of fuel cell commercialization is the ability to reduce costs to acceptable levels in both stationary and transport applications. It is increasingly recognized that the fuel processing subsystem can have a major impact on overall system costs, particularly as ongoing R and D efforts result in reduction of the basic cost structure of stacks which currently dominate system costs. The fuel processing subsystem for polymer electrolyte membrane fuel cell (PEMFC) technology, which is the focus of transport applications, includes the reformer, shift reactors, and means for CO reduction. In addition to low cost, transport applications require a fuel processor that is compact and can start rapidly. This paper describes the impact of factors such as fuel choice operating temperature, material selection, catalyst requirements, and controls on the cost of fuel processing systems. There are fuel processor technology paths which manufacturing cost analyses indicate are consistent with fuel processor subsystem costs of under $150/kW in stationary applications and $30/kW in transport applications. As such, the costs of mature fuel processing subsystem technologies should be consistent with their use in commercially viable fuel cell systems in both application categories. (orig.)

  13. Fuel manufacturing and utilization

    International Nuclear Information System (INIS)

    2005-01-01

    The efficient utilisation of nuclear fuel requires manufacturing facilities capable of making advanced fuel types, with appropriate quality control. Once made, the use of such fuels requires a proper understanding of their behaviour in the reactor environment, so that safe operation for the design life can be achieved. The International Atomic Energy Agency supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle. It provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection. The IAEA supports the development fuel modelling expertise in Member States, covering both normal operation and postulated and severe accident conditions. It provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation. The IAEA supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, it provides information and support research into the basic properties of fuel materials, including UO 2 , MOX and zirconium alloys. It further offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology

  14. High Temperature PEM Fuel Cells and Organic Fuels

    DEFF Research Database (Denmark)

    Vassiliev, Anton

    of the products. The observation of internal reforming was indirectly confirmed by electrochemical impedance spectroscopy, where the best fits were obtained when a Gerischer element describing preceding chemical reaction and diffusion was included in the equivalent circuit of a methanol/air operated cell...... evaporated liquid stream supply to either of the electrodes. A large number of MEAs with different component compositions have been prepared and tested in different conditions using the constructed setups to obtain a basic understanding of the nature of direct DME HT-PEM FC, to map the processes occurring...... inside the cells and to determine the lifetime. Additionally, comparison was made with methanol as fuel, which is the main competitor to DME in direct oxidation of organic fuels in fuel cells. For the reference, measurements have also been done with conventional hydrogen/air operation. All...

  15. Fuels and Combustion

    KAUST Repository

    Johansson, Bengt

    2016-08-17

    This chapter discusses the combustion processes and the link to the fuel properties that are suitable for them. It describes the basic three concepts, including spark ignition (SI) and compression ignition (CI), and homogeneous charge compression ignition (HCCI). The fuel used in a CI engine is vastly different from that in an SI engine. In an SI engine, the fuel should sustain high pressure and temperature without autoignition. Apart from the dominating SI and CI engines, it is also possible to operate with a type of combustion: autoignition. With HCCI, the fuel and air are fully premixed before combustion as in the SI engine, but combustion is started by the increased pressure and temperature during the compression stroke. Apart from the three combustion processes, there are also a few combined or intermediate concepts, such as Spark-Assisted Compression Ignition (SACI). Those concepts are discussed in terms of the requirements of fuel properties.

  16. Fuels and Combustion

    KAUST Repository

    Johansson, Bengt

    2016-01-01

    This chapter discusses the combustion processes and the link to the fuel properties that are suitable for them. It describes the basic three concepts, including spark ignition (SI) and compression ignition (CI), and homogeneous charge compression ignition (HCCI). The fuel used in a CI engine is vastly different from that in an SI engine. In an SI engine, the fuel should sustain high pressure and temperature without autoignition. Apart from the dominating SI and CI engines, it is also possible to operate with a type of combustion: autoignition. With HCCI, the fuel and air are fully premixed before combustion as in the SI engine, but combustion is started by the increased pressure and temperature during the compression stroke. Apart from the three combustion processes, there are also a few combined or intermediate concepts, such as Spark-Assisted Compression Ignition (SACI). Those concepts are discussed in terms of the requirements of fuel properties.

  17. Heat evaluation examination of fuel assembly

    International Nuclear Information System (INIS)

    Suto, Shinya; Nakabayashi, Hiroki; Yao, Kaoru

    2007-03-01

    The cooling examination was executed by using the simulated fuel assembly to obtain the basic data of the most effective cooling system in the lazer disassembling process of the spent fuel assembly of prototype fast breeder reactor 'Monju'. As a result, the following have been understood. (1) Before the laser disassembling (there is not any duct tube cutting), it is possible to cool enough by the amount of the wind of 20m 3 /h or more flowing from the handling head side. (2) After the laser disassembling begins (duct tube is cut), 1kW or more of the heat generation cannot be cooled by ventilation from the handling head side. (3) Cooling by the flow across fuel pin is required during lazer disassembling. The basic data of the cooling system was obtained from these examination results. However, for cooling across fuel pin during the laser disassembling, it is necessary to examine shape of the side cooling nozzle, spraying angle, and flow velocity at the nozzle exit, etc. enough. (author)

  18. Some recent contributions of basic nuclear science to nuclear waste transmutation

    International Nuclear Information System (INIS)

    Schapira, J.P.

    2001-01-01

    Nuclear waste transmutation aims at alleviating some long-term risks associated with actinides and with some long-lived fission products. Proposals of using accelerator driven system (ADS) to efficiently burn actinides in uranium free fuels have revitalized some basic researches in the field of nuclear and reactor physics. This is the case for high intensity accelerator in the ADS context and for the neutron source which relies to a large extent on basic nuclear physics related to spallation. There is also an experimental program called MUSE at Cadarache to study the sub-critical reactor physics with regard to its neutronics. A second area where basic research is involved is the measurement of new or more reliable neutron cross sections specific to transmutation and also to the thorium fuel cycle considered as a long-term option for ''clean'' energy production with reduced actinide production. This second area will possibly be covered by a new facility called n-TOF developed at CERN. (author)

  19. Some alternatives to the mixed oxide fuel cycle

    International Nuclear Information System (INIS)

    Deonigi, D.E.; Eschbach, E.A.; Goldsmith, S.; Pankaskie, P.J.; Rohrmann, C.A.; Widrig, R.D.

    1977-02-01

    While on initial examination each of the six fuel cycle concepts (tandem cycle, extended burnup, fuel rejuvenation, coprocessing, partial reprocessing, and thorium) described in the report may have some potential for improving safeguards, none of the six appears to have any other major or compelling advantages over the mixed oxide (MOX) fuel cycle. Compared to the MOX cycle, all but coprocessing appear to have major disadvantages, including severe cost penalties. Three of the concepts-tandem, extended burnup, and rejuvenation--share the basic problems of the throwaway cycle (GESMO Alternative 6): without reprocessing, high-level waste volumes and costs are substantially increased, and overall uranium utilization decreases for three reasons. First, the parasitic fission products left in the fuel absorb neutrons in later irradiation steps reducing the overall neutronic efficiencies of these cycles. Second, discarded fuel still has sufficient fissile values to warrant recycle. Third, perhaps most important, the plutonium needed for breeder start-up will not be available; without the breeder, uranium utilization would drop by about a factor of sixty. Two of the concepts--coprocessing and partial reprocessing--involve variations of the basic MOX fuel cycle's chemical reprocessing step to make plutonium diversion potentially more difficult. These concepts could be used with the MOX fuel cycle or in conjunction with the tandem, extended burnup and rejuvenation concepts to eliminate some of the problems with those cycles. But in so doing, the basic impetus for those cycles--elimination of reprocessing for safeguards purposes--no longer exists. Of all the concepts considered, only coprocessing--and particularly the ''master blend'' version--appears to have sufficient promise to warrant a more detailed study. The master blend concept could possibly make plutonium diversion more difficult with minimal impact on the reprocessing and MOX fuel fabrication operations

  20. Basic Considerations for Dry Storage of Spent Nuclear Fuels and Revisited CFD Thermal Analysis on the Concrete Cask

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Jae Soo [ACT Co. Ltd., Daejeon (Korea, Republic of); Park, Younwon; Song, Sub Lee [BEES Inc., Daejeon (Korea, Republic of); Kim, Hyeun Min [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    The integrity of storage facility and also of the spent nuclear fuel itself is considered very important. Storage casks can be located in a designated area on a site or in a designated storage building. A number of different designs for dry storage have been developed and used in different countries. Dry storage system was classified into two categories by IAEA. One is container including cask and silo, the other one is vault. However, there is various way of categorization for dry storage system. Dry silo and cask are usually classified separately, so the dry storage system can be classified into three different types. Furthermore, dry cask storage can be categorized into two types based on the type of the materials, concrete cask and metal cask. In this paper, the design characteristics of dry storage cask are introduced and computational fluid dynamics (CFD) based thermal analysis for concrete cask is revisited. Basic principles for dry storage cask design were described. Based on that, thermal analysis of concrete dry cask was introduced from the study of H. M. Kim et al. From the CFD calculation, the temperature of concrete wall was maintained under the safety criteria. From this fundamental analysis, further investigations are expected. For example, thermal analysis on the metal cask, thermal analysis on horizontally laid spent nuclear fuel assemblies for transportation concerns, and investigations on better performance of natural air circulation in dry cask can be promising candidates.

  1. Basic Considerations for Dry Storage of Spent Nuclear Fuels and Revisited CFD Thermal Analysis on the Concrete Cask

    International Nuclear Information System (INIS)

    Noh, Jae Soo; Park, Younwon; Song, Sub Lee; Kim, Hyeun Min

    2016-01-01

    The integrity of storage facility and also of the spent nuclear fuel itself is considered very important. Storage casks can be located in a designated area on a site or in a designated storage building. A number of different designs for dry storage have been developed and used in different countries. Dry storage system was classified into two categories by IAEA. One is container including cask and silo, the other one is vault. However, there is various way of categorization for dry storage system. Dry silo and cask are usually classified separately, so the dry storage system can be classified into three different types. Furthermore, dry cask storage can be categorized into two types based on the type of the materials, concrete cask and metal cask. In this paper, the design characteristics of dry storage cask are introduced and computational fluid dynamics (CFD) based thermal analysis for concrete cask is revisited. Basic principles for dry storage cask design were described. Based on that, thermal analysis of concrete dry cask was introduced from the study of H. M. Kim et al. From the CFD calculation, the temperature of concrete wall was maintained under the safety criteria. From this fundamental analysis, further investigations are expected. For example, thermal analysis on the metal cask, thermal analysis on horizontally laid spent nuclear fuel assemblies for transportation concerns, and investigations on better performance of natural air circulation in dry cask can be promising candidates

  2. R and D status and requirements for PIE in the fields of the HTGR fuel and the innovative basic research on High-Temperature Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Sawa, Kazuhiro; Tobita, Tsutomu; Sumita, Junya [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Ishihara, Masahiro; Hayashi, Kimio; Hoshiya, Taiji; Sekino, Hajime; Ooeda, Etsurou

    1999-09-01

    The High Temperature Engineering Test Reactor (HTTR), which is the first high temperature gas-cooled reactor (HTGR) in Japan, achieved its first criticality in November 1998 at the Oarai Research Establishment of the Japan Atomic Energy Research Institute (JAERI). In the field of HTGR fuel development, JAERI will proceed research and development (R and D) works by the following steps: (STEP-1) confirmation of irradiation performance of the first-loading fuel of the HTTR, (STEP-2) study on irradiation performance of high burnup SiC-coated fuel particle and (STEP-3) development of ZrC-coated fuel particle. Requirements for post-irradiation examination (PIE) are different for each R and D step. In STEP-1, firstly, hot cells will be prepared in the HTTR reactor building to handle spent fuels. In parallel, general equipments such as those for deconsolidation of fuel compacts and for handling coated fuel particles will be installed in the Hot Laboratory at Oarai. In STEP-2, precise PIE techniques, for example, Raman spectroscopy for measurement of stress on irradiated SiC layer, will be investigated. In STEP-3, new PIE techniques should be developed to investigate irradiation behavior of ZrC-coated particle. In the field of the innovative basic research on high-temperature engineering, some preliminary tests have been made on the research areas of (1) new materials development, (2) fusion technology, (3) radiation chemistry and (4) high-temperature in-core instrumentation. Requirements for PIE are under investigation, in particular in the field of the new materials development. Besides more general apparatuses including transmission electron microscopy (TEM), some special apparatuses such as an electron spin resonance (ESR) spectrometer, a specific resistance/Hall coefficient measuring system and a differential scanning calorimeter (DSC) are planned to install in the Hot Laboratory at Oarai. Acquisition of advanced knowledge on the irradiation behavior is expected in

  3. The DUPIC alternative for backend fuel cycle

    International Nuclear Information System (INIS)

    Lee, J.S.; Choi, J.W.; Park, H.S.; Boczar, P.; Sullivan, J.; Gadsby, R.D.

    1997-01-01

    From the early nineties, a research programme, called DUPIC (Direct Use of Spent PWR Fuel in CANDU) has been undertaken in an international exercise involving Korea, Canada, the U.S. and later the IAEA. The basic idea of this fuel cycle alternative is that the spent fuel from LWR contains enough fissile remnant to be burnt again in CANDUs thanks to its excellent neutron economy. A systematic R and D plan has now gained a full momentum to verify experimentally the DUPIC fuel cycle concept. 4 refs

  4. Fuel Cell/Electrochemical Cell Voltage Monitor

    Science.gov (United States)

    Vasquez, Arturo

    2012-01-01

    A concept has been developed for a new fuel cell individual-cell-voltage monitor that can be directly connected to a multi-cell fuel cell stack for direct substack power provisioning. It can also provide voltage isolation for applications in high-voltage fuel cell stacks. The technology consists of basic modules, each with an 8- to 16-cell input electrical measurement connection port. For each basic module, a power input connection would be provided for direct connection to a sub-stack of fuel cells in series within the larger stack. This power connection would allow for module power to be available in the range of 9-15 volts DC. The relatively low voltage differences that the module would encounter from the input electrical measurement connection port, coupled with the fact that the module's operating power is supplied by the same substack voltage input (and so will be at similar voltage), provides for elimination of high-commonmode voltage issues within each module. Within each module, there would be options for analog-to-digital conversion and data transfer schemes. Each module would also include a data-output/communication port. Each of these ports would be required to be either non-electrical (e.g., optically isolated) or electrically isolated. This is necessary to account for the fact that the plurality of modules attached to the stack will normally be at a range of voltages approaching the full range of the fuel cell stack operating voltages. A communications/ data bus could interface with the several basic modules. Options have been identified for command inputs from the spacecraft vehicle controller, and for output-status/data feeds to the vehicle.

  5. WWER-440 fuel cycles possibilities using improved fuel assemblies design

    International Nuclear Information System (INIS)

    Mikolas, P.; Svarny, J.

    2008-01-01

    Practically five years cycle has been achieved in the last years at NPP Dukovany. There are two principal means how it could be achieved. First, it is necessary to use fuel assemblies with higher fuel enrichment and second, to use fuel loading with very low leakage. Both these conditions are fulfilled at NPP Dukovany at this time. It is known, that the fuel cycle economy can be improved by increasing the fuel residence time in the core up to six years. There are at least two ways how this goal could be achieved. The simplest way is to increase enrichment in fuel. There exists a limit, which is 5.0 w % of 235 U. Taking into account some uncertainty, the calculation maximum is 4.95 w % of 235 U. The second way is to change fuel assembly design. There are several possibilities, which seem to be suitable from the neutron - physical point of view. The first one is higher mass content of uranium in a fuel assembly. The next possibility is to enlarge pin pitch. The last possibility is to 'omit' FA shroud. This is practically unrealistic; anyway, some other structural parts must be introduced. The basic neutron physical characteristics of these cycles for up-rated power are presented showing that the possibilities of fuel assemblies with this improved design in enlargement of fuel cycles are very promising. In the end, on the basis of neutron physical characteristics and necessary economical input parameters, a preliminary evaluation of economic contribution of proposals of advanced fuel assemblies on fuel cycle economy is presented (Authors)

  6. Introduction to the study of the treatment of spent nuclear fuel

    International Nuclear Information System (INIS)

    Floh, B.; Araujo, J.A. de; Matsuda, H.T.

    1975-01-01

    An introduction is made to the study of the treatment of spent nuclear fuels. Main topics discussed are: basic information, volatilization processes, treatment of thorium based fuels (Thorex process), analytical chemistry of spent nuclear fuel and design of industrial facilities

  7. The role of spent fuel test facilities in the fuel cycle strategy

    International Nuclear Information System (INIS)

    Huang, S. T.; Gross, D. L.; Snyder, N. W.; Woods, W. D.

    1988-01-01

    Disposal of commercial spent nuclear fuels in the major industrialized countries may be categorized into two broad approaches: a once-through policy which will dispose of spent fuels and recycle fissile materials. Within reprocess spent fuels and recycle fissile materials. Within each policy, various technical, licensing, institutional and public issues exist. These issues tend to complicate the formulation of an effective and acceptable fuel cycle strategy which will meet various cost, schedule, and legislative constraints. This paper examines overall fuel cycle strategies from the viewpoint of these underlying technical issues and assesses the roles of spent fuel test facilities in the overall fuel cycles steps. Basic functions of such test facilities are also discussed. The main emphasis is placed on the once-through policy although the reprocessing / recycle policy is also discussed. Benefits of utilizing test facilities in the fuel cycle strategies are explored. The results indicate that substantial benefits may be obtained in terms of minimizing programmatic risks, increasing public confidence, and more effective utilization of overall budgetary resources by structuring and highlighting the test facilities as an important element in the overall strategy

  8. Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

  9. The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology

    International Nuclear Information System (INIS)

    Bobrov, E.; Alekseev, P.; Chibinyaev, A.; Teplov, P.; Dudnikov, A.

    2016-01-01

    REMIX (Regenerated Mixture) fuel is produced directly from a non-separated mix of recycled uranium and plutonium from reprocessed used fuel and the fabrication technology of such fuel is called REMIX-technology. This paper shows basic features of different fuel assembly (FA) application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water-fuel ratio in the VVER FA affects the fuel characteristics produced by REMIX technology during multiple recycling. It is shown that for for the traditional REMIX-fuel it does not make sense to change anything in the design of VVER FA, because there are no advantages in the fuel feed consumption. The natural uranium economy by the fifth cycle reached about 29%. In the case of the REMIX fuel based on uranium-plutonium from SNF MOX fuel, it would be appropriate to use fuel assemblies with a water-fuel ratio of 1.5

  10. Burnup credit demands for spent fuel management in Ukraine

    International Nuclear Information System (INIS)

    Medun, V.

    2001-01-01

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

  11. New Concept of Designing Composite Fuel for Fast Reactors with Closing Fuel Cycle

    International Nuclear Information System (INIS)

    Savchenko, A.; Vatulin, A.; Uferov, O.; Kulakov, G.; Sorokin, V.

    2013-01-01

    For fast reactors a novel type of promising composite U-PuO2 fuel is proposed which is based on dispersion fuel elements. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. Novel fuel features higher characteristics in comparison to metallic or MOX fuel its fabrication technology is readily accomplished and is environmentally clean. A possibility is demonstrated of fabricating coated steel claddings to protect from interaction with fuel and fission products when use standard rod type MOX or metallic U-Pu-Zr fuel. Novel approach to reprocessing of composite fuel is demonstrated, which allows to separate uranium from burnt plutonium as well as the newly generated fissile plutonium from burnt one without chemical processes, which simplifies the closing of the nuclear fuel cycle. Novel composite fuel combines the advantages of metallic and ceramic types of fuel and has high uranium density that allows also to implicate it in BREST types reactor with conversion ratio more than 1. Peculiarities of closing nuclear cycle with composite fuel are demonstrated that allows more effective re-usage of generated Pu as well as, minimizing r/a wastes by incineration of MA in specially developed IMF design

  12. Energy, equity and the future of the fuel poor

    International Nuclear Information System (INIS)

    Roberts, Simon

    2008-01-01

    A warm and adequately-lit home is considered a basic need, together with access to energy-consuming appliances ranging from a fridge to a TV. An underlying tenet of sustainable energy is that such basic needs should be affordably met. Yet low incomes, energy-inefficient housing and appliances and high energy costs mean that roughly 10 per cent of UK households, many of them elderly or with young children, fail to attain this basic standard. These households, which would need to spend more than 10 per cent of their income to attain adequate energy services, are officially defined as 'fuel poor'. Their cold, poorly equipped homes lead to chronic cold-related health conditions, exacerbate social isolation, and may undermine educational achievement. In addition, rural areas have a disproportionately high incidence of fuel poverty. This Review examines the current distribution of energy consumption, its social impacts, and the opportunities to address fuel poverty through improvements to the housing stock. It will then consider potential future developments

  13. How Canada has controlled the spent fuel storage problem

    International Nuclear Information System (INIS)

    Mosey, D.

    1985-01-01

    A report on the irradiated fuel storage workshop held in Toronto in October 1984. In particular Canada's attitude to spent fuel is examined. The basic fuel cycle has been envisaged as running from mining and refining, through interim storage to final geologic disposal, with reprocessing as an option to be considered when it looks economically attractive. (U.K.)

  14. Wildland fire emissions, carbon, and climate: Modeling fuel consumption

    Science.gov (United States)

    Roger D. Ottmar

    2014-01-01

    Fuel consumption specifies the amount of vegetative biomass consumed during wildland fire. It is a two-stage process of pyrolysis and combustion that occurs simultaneously and at different rates depending on the characteristics and condition of the fuel, weather, topography, and in the case of prescribed fire, ignition rate and pattern. Fuel consumption is the basic...

  15. HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1977. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.

  16. Future plant of basic research for nuclear energy by university researchers

    International Nuclear Information System (INIS)

    Shibata, Toshikazu

    1984-01-01

    National Committee for Nuclear Energy Research, Japan Science Council has completed a future plan for basic nuclear energy research by university researchers. The JSC has recommended the promotion of basic research for nuclear energy based on the plan in 1983. The future plan consists of four main research fields, namely, (1) improvements of reactor safety, (2) down stream, (3) thorium fuel reactors, and (4) applications of research reactor and radioisotopes. (author)

  17. Reactor fuel performance data file, 1985 edition

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Fujita, Misao; Watanabe, Kohji.

    1986-07-01

    In safety evaluation and integrity studies of reactor fuel, data on fuel performance are the most basic materials. The Fuel Reliability Laboratory No.1 has obtained the fuel performance data by joining in some international programs to study the safety and integrity of fuel. Those data have only used for the studies in the above two fields. However, if the data are rearranged and compiled in a easily usable form, they can be utilized in other field of studies. Then, a 'data file' on fuel performance is beeing compiled by adding data from open literatures to those obtained in international programs. The present report is prepared on the basis of the data file compiled by March in 1986. (author)

  18. Concerning 1991 basic plan for atomic energy development and application (subjected to examination)

    International Nuclear Information System (INIS)

    1990-01-01

    The prime minister developed a draft 1991 Basic Plan for Atomic Energy Development and Application and sent it to the Nuclear Safety Commission for examination. The Commission started the examination at its 14th meeting. The report outlines results of the examination. A Basic Plan is developed each year to promote efforts at atomic energy development and application systematically and efficiently. In particular, it identifies specific activities required to realize the basic policies shown in the Long Term Program for Atomic Energy Development and Application. In the present report, activities required for improving the safety measures in general are described first, with special emphasis placed on the improvement in nuclear safety regulations and promotion of nuclear safety research. Activities required for promoting nuclear power generation are then outlined. It also insists that the nuclear fuel cycle should be established by promoting measures for uranium resources, uranium enrichment, spent fuel enrichment, and radioactive waste disposal. Other required efforts include the development of improved power reactors, implementation of major projects, and development of basic technology. (N.K.)

  19. Metal Matrix Microencapsulated Fuel Technology for LWR Applications

    International Nuclear Information System (INIS)

    Terrani, Kurt A.; Bell, Gary L.; Kiggans, Jim; Snead, Lance Lewis

    2012-01-01

    An overview of the metal matrix microencapsulated (M3) fuel concept for the specific LWR application has been provided. Basic fuel properties and characteristics that aim to improve operational reliability, enlarge performance envelope, and enhance safety margins under design-basis accident scenarios are summarized. Fabrication of M3 rodlets with various coated fuel particles over a temperature range of 800-1300 C is discussed. Results from preliminary irradiation testing of LWR M3 rodlets with surrogate coated fuel particles are also reported.

  20. Advanced fuel cycle on the basis of pyroelectrochemical process for irradiated fuel reprocessing and vibropacking technology

    International Nuclear Information System (INIS)

    Mayorshin, A.A.; Skiba, O.V.; Tsykanov, V.A.; Golovanov, V.N.; Bychkov, A.V.; Kisly, V.A.; Bobrov, D.A.

    2000-01-01

    For advanced nuclear fuel cycle in SSC RIAR there is developed the pyroelectrochemical process to reprocess irradiated fuel and produce granulated oxide fuel UO 2 , PuO 2 or (U,Pu)O 2 from chloride melts. The basic technological stage is the extraction of oxides as a crystal product with the methods either of the electrolysis (UO 2 and UO 2 -PuO 2 ) or of the precipitating crystalIization (PuO 2 ). After treating the granulated fuel is ready for direct use to manufacture vibropacking fuel pins. Electrochemical model for (U,Pu)O 2 coprecipitation is described. There are new processes being developed: electroprecipitation of mixed oxides - (U,Np)O 2 , (U,Pu,Np)O 2 , (U,Am)O 2 and (U,Pu,Am)O 2 . Pyroelectrochemical production of mixed actinide oxides is used both for reprocessing spent fuel and for producing actinide fuel. Both the efficiency of pyroelectrochemical methods application for reprocessing nuclear fuel and of vibropac technology for plutonium recovery are estimated. (author)

  1. Studies and research concerning BNFP: LWR spent fuel storage

    International Nuclear Information System (INIS)

    Shallo, F.A.

    1978-08-01

    This report describes potential spent fuel storage expansion programs using the Barnwell Nuclear Fuel Plant--Fuel Receiving and Storage Station (BNFP-FRSS) as a model. Three basic storage arrangements are evaluated with cost and schedule estimates being provided for each configuration. A general description of the existing facility is included with emphasis on the technical and equipment requirements which would be necessary to achieve increased spent fuel storage capacity at BNFP-FRSS

  2. CANDU-PHW fuel management

    International Nuclear Information System (INIS)

    Frescura, G.M.; Wight, A.L.

    1982-01-01

    This report covers the material presented in a series of six lectures at the Winter College on Nuclear Physics and Reactors held at the International Centre for Theoretical Physics in Trieste, Italy, Jan 22 - March 28, 1980. The report deals with fuel management in natural uranium fuelled CANDU-PHW reactors. Assuming that the reader has a basic knowledge of CANDU core physics, some of the reactor systems which are more closely related to fuelling are described. This is followed by a discussion of the methods used to calculate the power distribution and perform fuel management analyses for the equilibrium core. A brief description of some computer codes used in fuel management is given, together with an overview of the calculations required to provide parameters for core design and support the accident analysis. Fuel scheduling during approach to equilibrium and equilibrium is discussed. Fuel management during actual reactor operation is discussed with a review of the operating experience for some of the Ontario Hydro CANDU reactors. (author)

  3. Recent prospects of MOX fuel and strategy about nuclear fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1991-04-01

    It is clearly described what is the preliminary adequate strategic concern for different nuclear power countries under different nuclear power development conditions. It is also stressed on the basic situation of the design technology, manufacture technology, operation experiences and quantitative economic analysis for MOX fuel application since fast breed reactor commercialization has been delayed. The author specially proposed that in a short term China should adopt an intermediate storage strategy matched with the construction of a pilot reprocessing plant to prepare the technical basis for commercialized reprocessing plant later on and to follow the development of MOX fuel technology

  4. Nuclear-fuel-cycle optimization: methods and modelling techniques

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1982-01-01

    This book present methods applicable to analyzing fuel-cycle logistics and optimization as well as in evaluating the economics of different reactor strategies. After an introduction to the phases of a fuel cycle, uranium cost trends are assessed in a global perspective. Subsequent chapters deal with the fuel-cycle problems faced by a power utility. The fuel-cycle models cover the entire cycle from the supply of uranium to the disposition of spent fuel. The chapter headings are: Nuclear Fuel Cycle, Uranium Supply and Demand, Basic Model of the LWR (light water reactor) Fuel Cycle, Resolution of Uncertainties, Assessment of Proliferation Risks, Multigoal Optimization, Generalized Fuel-Cycle Models, Reactor Strategy Calculations, and Interface with Energy Strategies. 47 references, 34 figures, 25 tables

  5. Computer code TOBUNRAD for PWR fuel bundle heat-up calculations

    International Nuclear Information System (INIS)

    Shimooke, Takanori; Yoshida, Kazuo

    1979-05-01

    The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)

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

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

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

  7. Trends in fuel reprocessing safety research

    International Nuclear Information System (INIS)

    Tsujino, Takeshi

    1981-01-01

    With the operation of a fuel reprocessing plant in the Power Reactor and Nuclear Fuel Development Corporation (PNC) and the plan for a second fuel reprocessing plant, the research on fuel reprocessing safety, along with the reprocessing technology itself, has become increasingly important. As compared with the case of LWR power plants, the safety research in this field still lags behind. In the safety of fuel reprocessing, there are the aspects of keeping radiation exposure as low as possible in both personnel and local people, the high reliability of the plant operation and the securing of public safety in accidents. Safety research is then required to establish the safety standards and to raise the rate of plant operation associated with safety. The following matters are described: basic ideas for the safety design, safety features in fuel reprocessing, safety guideline and standards, and safety research for fuel reprocessing. (J.P.N.)

  8. Basic data generation and pressure loss coefficient evaluation for HANARO core thermal-hydraulic analyses

    International Nuclear Information System (INIS)

    Chae, Hee Taek; Lee, Kye Hong

    1999-06-01

    MATRA-h, a HANARO subchannel analysis computer code, is used to evaluate thermal margin of the HANARO fuel. It's capability includes the assessments of CHF, ONB margin, and fuel temperature. In this report, basic input data and core design parameters required to perform the subchannel analysis with MATRA-h code are collected. These data include the subchannel geometric data, thermal-hydraulic correlations, empirical constants and material properties. The friction and form loss coefficients of the fuel assemblies were determined based on the results of the pressure drop test. At the same time, different form loss coefficients at the end plates and spacers are evaluated for various subchannels. The adequate correlations are applied to the evaluation of the form loss coefficients for various subchannels, which are corrected by measured values in order to have a same pressure drop at each flow channel. These basic input data and design parameters described in this report will be applied usefully to evaluate the thermal margin of the HANARO fuel. (author). 11 refs., 13 tabs., 11 figs

  9. PWR and WWER fuel performance. A comparison of major characteristics

    International Nuclear Information System (INIS)

    Weidinger, H.

    2006-01-01

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  10. Fuel utilization experience in Bohunice NPP and regulatory requirements for implementation of progressive fuel management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Patenyi, V [Nuclear Regulatory Authority, Bratislava (Slovakia); Darilek, P; Majercik, J [Vyskumny Ustav Jadrovych Elektrarni, Trnava (Slovakia)

    1994-12-31

    The experience gained in fuel utilization and the basic requirements for fuel licensing in the Slovak NPPs is described. The original project of WWER-440 reactors supposes 3-year fuel cycle with cycle length of about 320 full power days (FPD). Since 1984 it was reduced to 290 FPD. Based on the experience of other countries, a 4-year fuel cycle utilization started in 1987. It is illustrated with data from the Bohunice NPP units. Among 504 fuel assemblies left for the fourth burnup cycle no leakage was observed. The mean burnup achieved in the different units varied from 33.1 to 38.5 Mwd/kg U. The new fuel assemblies used are different from the recent ones in construction, thermohydraulics, water-uranium ratio, enrichment and material design. To meet the safety criteria, regulatory requirements for exploitation of new fuel in WWER-440 were formulated by the Nuclear Regulatory Authority of Slovak Republic. 1 tab., 5 refs.

  11. Detailed analysis of uranium silicide dispersion fuel swelling

    International Nuclear Information System (INIS)

    Hofmann, G.L.; Ryu, Woo-Seog

    1991-01-01

    Swelling of U 3 Si and U 3 Si 2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and micro structural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide dispersion fuel. (orig.)

  12. Basic design of radiation-resistant LVDTs: Linear Variable Differential Transformer

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, J. M.; Park, S. J.; Kang, Y. H. (and others)

    2008-02-15

    A LVDT(Linear Variable Differential Transformer) for measuring the pressure level was used to measure the pressure of a nuclear fuel rod during the neutron irradiation test in a research reactor. A LVDT for measuring the elongation was also used to measure the elongation of nuclear fuels, and the creep and fatigue of materials during a neutron irradiation test in a research reactor. In this report, the basic design of two radiation-resistant LVDTs for measuring the pressure level and elongation are described. These LVDTs are used a under radiation environment such as a research reactor. In the basic design step, we analyzed the domestic and foreign technical status for radiation-resistant LVDTs, made part and assembly drawings and established simple procedures for their assembling. Only a few companies in the world can produce radiation-resistant LVDTs. Not only these are extremely expensive, but the prices are continuously rising. Also, it takes a long time to procure a LVDT, as it can only be bought about by an order-production. The localization of radiation-resistant LVDTs is necessary in order to provide them quickly and at a low cost. These radiation-resistant LVDTs will be used at neutron irradiation devices such as instrumented fuel capsules, special purpose capsules and a fuel test loop in research reactors. We expect that the use of neutron irradiation tests will be revitalized by the localization of radiation-resistant LVDTs.

  13. Experimental research of fuel element reliability

    International Nuclear Information System (INIS)

    Cech, B.; Novak, J.; Chamrad, B.

    1980-01-01

    The rate and extent of the damage of the can integrity for fission products is the basic criterion of reliability. The extent of damage is measurable by the fission product leakage into the reactor coolant circuit. An analysis is made of the causes of the fuel element can damage and a model is proposed for testing fuel element reliability. Special experiments should be carried out to assess partial processes, such as heat transfer and fuel element surface temperature, fission gas liberation and pressure changes inside the element, corrosion weakening of the can wall, can deformation as a result of mechanical interactions. The irradiation probe for reliability testing of fuel elements is described. (M.S.)

  14. Manufacturing experience and perspectives of WWER nuclear fuel development

    International Nuclear Information System (INIS)

    Aksenov, P.; Kolosovskiy, V.

    2011-01-01

    The purposes of new shroudless working fuel assembly (PK-3) development, basic design peculiarities of working fuel assembly (PK-3) and the results of PK-3 implementation are presented in this paper. Values of 440.19.000-02 working fuel assembly with debris filter Burnup at Kola NPP unit 2 are given. The main issues settled in the course of TVSA-T implementation like: The development of the design and fabrication method of mixing grids; The development of the design and fabrication method of basic assemblies and components of TVSA-T, including fuel rods of new generation; and The obtainment of specified pellet microstructure with average grain size more than 25μm are listed. The development of the design and fabrication method of removable uprated headpiece of shortened length as well as the development of the design and fabrication method of a tailpiece equipped with a debris filter are also illustrated

  15. Conceptual structure design of experimental facility for advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    Joo, J. S.; Koo, J. H.; Jung, W. M.; Jo, I. J.; Kook, D. H.; Yoo, K. S.

    2003-01-01

    A study on the advanced spent fuel conditioning process (ACP) is carring out for the effective management of spent fuels of domestic nuclear power plants. This study presents basic shielding design, modification of IMEF's reserve hot cell facility which reserved for future usage, conceptual and structural architecture design of ACP hot cell and its contents, etc. considering the characteristics of ACP. The results of this study will be used for the basic and detail design of ACP demonstration facility, and utilized as basic data for the safety evaluation as essential data for the licensing of the ACP facility

  16. Basic Data on Biogas

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-07-01

    Renewable gases such as biogas and biomethane are considered as key energy carrier when the society is replacing fossil fuels with renewable alternatives. In Sweden, almost 80 % of the fossil fuels are used in the transport sector. Therefore, the focus in Sweden has been to use the produced biogas in this sector as vehicle gas. Basic Data on Biogas contains an overview of production, utilisation, climate effects etc. of biogas from a Swedish perspective. The purpose is to give an easy overview of the current situation in Sweden for politicians, decision makers and interested public. 1.4 TWh of biogas is produced annually in Sweden at approximately 230 facilities. The 135 wastewater treatment plants that produce biogas contribute with around half of the production. In order to reduce the sludge volume, biogas has been produced at wastewater treatment plants for decades. New biogas plants are mainly co-digestion plants and farm plants. The land filling of organic waste has been banned since 2005, thus the biogas produced in landfills is decreasing.

  17. QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1968 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J.; de Halas, D. R.; Nightingale, R. E.; Worlton, D. C.

    1968-06-01

    Progress is reported in these areas: nuclear graphite; fuel development for gas-cooled reactors; HTGR graphite studies; nuclear ceramics; fast-reactor nitrides research; non-destructive testing; metallic fuels; basic swelling studies; ATR gas and water loop operation and maintenance; reactor fuels and materials; fast reactor dosimetry and damage analysis; and irradiation damage to reactor metals.

  18. Modelling the effects of transport policy levers on fuel efficiency and national fuel consumption

    International Nuclear Information System (INIS)

    Kirby, H.R.; Hutton, B.; McQuaid, R.W.; Napier Univ., Edinburgh; Raeside, R.; Napier Univ., Edinburgh; Zhang, Xiayoan; Napier Univ., Edinburgh

    2000-01-01

    The paper provides an overview of the main features of a Vehicle Market Model (VMM) which estimates changes to vehicle stock/kilometrage, fuel consumed and CO 2 emitted. It is disaggregated into four basic vehicle types. The model includes: the trends in fuel consumption of new cars, including the role of fuel price: a sub-model to estimate the fuel consumption of vehicles on roads characterised by user-defined driving cycle regimes; procedures that reflect distribution of traffic across different area/road types; and the ability to vary the speed (or driving cycle) from one year to another, or as a result of traffic growth. The most significant variable influencing fuel consumption of vehicles was consumption in the previous year, followed by dummy variables related to engine size. the time trend (a proxy for technological improvements), and then fuel price. Indeed the effect of fuel price on car fuel efficiency was observed to be insignificant (at the 95% level) in two of the three versions of the model, and the size of fuel price term was also the smallest. This suggests that the effectiveness of using fuel prices as a direct policy tool to reduce fuel consumption may he limited. Fuel prices may have significant indirect impacts (such as influencing people to purchase more fuel efficient cars and vehicle manufacturers to invest in developing fuel efficient technology) as may other factors such as the threat of legislation. (Author)

  19. Limitations of Commercializing Fuel Cell Technologies

    Science.gov (United States)

    Nordin, Normayati

    2010-06-01

    Fuel cell is the technology that, nowadays, is deemed having a great potential to be used in supplying energy. Basically, fuel cells can be categorized particularly by the kind of employed electrolyte. Several fuel cells types which are currently identified having huge potential to be utilized, namely, Solid Oxide Fuel Cells (SOFC), Molten Carbonate Fuel Cells (MCFC), Alkaline Fuel Cells (AFC), Phosphoric Acid Fuel Cells (PAFC), Polymer Electron Membrane Fuel Cell (PEMFC), Direct Methanol Fuel Cells (DMFC) and Regenerative Fuel Cells (RFC). In general, each of these fuel cells types has their own characteristics and specifications which assign the capability and suitability of them to be utilized for any particular applications. Stationary power generations and transport applications are the two most significant applications currently aimed for the fuel cell market. It is generally accepted that there are lots of advantages if fuel cells can be excessively commercialized primarily in context of environmental concerns and energy security. Nevertheless, this is a demanding task to be accomplished, as there is some gap in fuel cells technology itself which needs a major enhancement. It can be concluded, from the previous study, cost, durability and performance are identified as the main limitations to be firstly overcome in enabling fuel cells technology become viable for the market.

  20. Qualification status of LEU [low enriched uranium] fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.

    1987-01-01

    Sufficient data has been obtained from tests of high-density, low-enriched fuels for research and test reactors to declare them qualified for use. These fuels include UZrH x (TRIGA fuel) and UO 2 (SPERT fuel) for rod-type reactors and UAl x , U 3 O 8 , U 3 Si 2 , and U 3 Si dispersed in aluminium for plate-type reactors. Except for U 3 Si, the allowable fission density for LEU applications is limited only by the available 235 U. Several reactors are now using these fuels, and additional conversions are in progress. The basic performance characteristics and limits, if any, of the qualified low-enriched (and medium-enriched) fuels are discussed. Continuing and planned work to qualify additional fuels is also discussed. (Author)

  1. The Atomic energy basic law

    International Nuclear Information System (INIS)

    1979-01-01

    The law aims to secure future energy resources, push forward progress of science and advancement of industry for welfare of the mankind and higher standard of national life by helping research, development and utilization of atomic power. Research, development and utilization of atomic power shall be limited to the peaceful purpose with emphasis laid on safety and carried on independently under democratic administration. Basic concepts and terms are defined, such as: atomic power; nuclear fuel material; nuclear raw material; reactor and radiation. The Atomic Energy Commission and the Atomic Energy Safety Commission shall be set up at the Prime Minister's Office deliberately to realize national policy of research, development and utilization of atomic power and manage democratic administration for atomic energy. The Atomic Energy Commission shall plan, consider and decide matters concerning research, development and utilization of atomic energy. The Atomic Energy Safety Commission shall plan, consider and decide issues particularly concerning safety securing among such matters. The Atomic Energy Research Institute shall be founded under the governmental supervision to perform research, experiment and other necessary affairs for development of atomic energy. The Power Reactor and Nuclear Fuel Development Corporation shall be established likewise to develop fast breeding reactor, advanced thermal reactor and nuclear fuel materials. Development of radioactive minerals, control of nuclear fuel materials and reactors and measures for patent and invention concerning atomic energy, etc. are stipulated respectively. (Okada, K.)

  2. International fuel cycle centres offer large economics and easier financing

    International Nuclear Information System (INIS)

    Smith, D.

    1977-01-01

    The summary report of the IAEA study project on multi-national regional nuclear fuel cycle indicates that for facilities of reasonable size such projects offer very decisive advantages in fuel cycle costs and resource availability over national facilities in general, and more markedly over the other alternative of the open ended, non-recycle fuel route. The economic evaluation of alternative fuel cycle strategies, one of the basic studies summarised in the report, is considered. (author)

  3. Structural analysis of advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    Gu, J. H.; Jung, W. M.; Jo, I. J.; Gug, D. H.; Yoo, K. S.

    2003-01-01

    An advanced spent fuel conditioning process (ACP) is developing for the safe and effective management of spent fuels which arising from the domestic nuclear power plants. And its demonstration facility is under design. This facility will be prepared by modifying IMEF's reserve hot cell facility which reserved for future usage by considering the characteristics of ACP. This study presents a basic structural architecture design and analysis results of ACP hot cell including modification of the IMEF. The results of this study will be used for the detail design of ACP demonstration facility, and utilized as basic data for the licensing of the ACP facility

  4. Irradiation and performance evaluation of DUPIC fuel

    International Nuclear Information System (INIS)

    Bae, Ki Kwang; Yang, M. S.; Song, K. C.

    2000-05-01

    The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis

  5. Irradiation and performance evaluation of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Ki Kwang; Yang, M S; Song, K C [and others

    2000-05-01

    The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis.

  6. Dissolution studies of spent nuclear fuels

    International Nuclear Information System (INIS)

    1991-02-01

    To obtain quantitative data on the dissolution of high burnup spent nuclear fuel, dissolution study have been carried out at the Department of Chemistry, JAERI, from 1984 under the contract with STA entitled 'Reprocessing Test Study of High Burnup Fuel'. In this study PWR spent fuels of 8,400 to 36,100 MWd/t in averaged burnup were dissolved and the chemical composition and distribution of radioactive nuclides were measured for insoluble residue, cladding material (hull), off-gas and dissolved solution. With these analyses basic data concerning the dissolution and clarification process in the reprocessing plant were accumulated. (author)

  7. Detailed analysis of uranium silicide dispersion fuel swelling

    International Nuclear Information System (INIS)

    Hofman, G.L.; Ryu, Woo-Seog.

    1989-01-01

    Swelling of U 3 Si and U 3 Si 2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and microstructural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide disperson fuel. 5 refs., 10 figs

  8. Future Transient Testing of Advanced Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack

    2009-09-01

    The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by

  9. Future Transient Testing of Advanced Fuels

    International Nuclear Information System (INIS)

    Carmack, Jon

    2009-01-01

    The transient in-reactor fuels testing workshop was held on May 4-5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat energie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric - Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by the

  10. Overview of the spent fuel management policy in Finland

    International Nuclear Information System (INIS)

    Manninen, Jussi

    1985-01-01

    The basic factors affecting the spent fuel management policy are highlighted: small size of the nuclear programme in the worldwide scale, no recycling of plutonium envisaged, no governmental organizations for back-end operations foreseen. The prinsiple objective of the policy permanent disposal of high-level wastes irrevocably outside the domestic territory, and the limited success in its implementation are discussed. The preparations of the implementation of the back-up alternative, direct disposal of spent fuel in the Finnish bedrock are described. The basic philosophy behind the system of funding the future waste management costs is clarified. (author)

  11. Fuel cycle management in Finland

    International Nuclear Information System (INIS)

    Vaeyrynen, H.; Mikkola, I.

    1987-01-01

    Both Finnish utilities producing nuclear power - Imatran Voima Oy (IVO) and Teollisuuden Voima Oy (Industrial Power Co. Ltd, TVO) - have created efficient fuel cycle management systems. The systems however differ in almost all respects. The reason is that the principal supplier for IVO is the Soviet Union and for TVO is Sweden. A common feature of both systems at the front end of the cycle is the building of stockpiles in order to provide for interruptions in fuel deliveries. Quality assurance supervision at the fuel factory for IVO is regulated by the Soviet Chamber of Commerce and Industry and a final control is made in Finland. The in-core fuel management is done by IVO using codes developed in Finland. The whole IVO fuel cycle is basically a leasing arrangement. The spent fuel is returned to the USSR after five years cooling. TVO carries out the in-core fuel management using a computer code system supplied by Asea-Atom. TVO is responsable for the back end of the cycle and makes preparations for the final disposal of the spent fuel in Finland. 6 refs., 2 figs

  12. Oxygen Chemical Diffusion Coefficients of (Pu,Am)O2 Fuels

    International Nuclear Information System (INIS)

    Watanabe, M.; Kato, M.; Matsumoto, T.

    2015-01-01

    Minor actinide (MA)-bearing MOX fuels have been developed as candidate fuels which are used in fast neutron spectrum cores such as sodium-cooled fast reactor (SFR) cores and experimental accelerator driven system (ADS) cores. Americium (Am) which is one of the MA elements significantly affects basic properties. It is known that Am content causes oxygen potential to increase and that influences irradiation behaviour such as fuel-cladding chemical interaction (FCCI) and chemical state of fission products. However, the effects of Am content on changes of basic properties are not clear. In this work, the oxygen chemical diffusion coefficients were calculated from measured data and the relationship between oxygen diffusion and oxygen potential of (Pu,Am)O 2-x was discussed. (authors)

  13. An introduction to the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Leuze, R.E.

    1986-01-01

    This overview of the nuclear fuel cycle is divided into three parts. First, is a brief discussion of the basic principles of how nuclear reactors work;second, is a look at the major types of nuclear reactors being used and world-wide nuclear capacity;and third, is an overview of the nuclear fuel cycle and the present industrial capability in the US. 34 figs., 10 tabs

  14. Ore potential of basic rocks in Finland

    International Nuclear Information System (INIS)

    Reino, J.; Ekberg, M.; Heinonen, P.; Karppanen, T.; Hakapaeae, A.; Sandberg, E.

    1993-02-01

    The report is associated with a study programme on basic rocks, which has the aim to complement the preliminary site investigations on repository for TVO's (Teollisuuden Voima Oy) spent nuclear fuel. The report comprises a mining enterprise's view of the ore potential of basic plutonic rocks in Finland. The ores associated with basic plutonic rocks are globally known and constitute a significant share of the global mining industry. The ores comprise chromium, vanadium-titanium-iron, nickel-copper and platinum group element ores. The resources of the metals in question and their mining industry are examined globally. A review of the use of these metals in the industry is presented as well. General factors affecting the mining industry, such as metal prices, political conjunctures, transport facilities, environmental requirements and raw material sources for the Finnish smelters have been observed from the point of view of their future effect on exploration activity and industrial development in Finland. Information on ores and mineralizations associated with Finnish basic rocks have been compiled in the report. The file comprises 4 chromium occurrences, 8 vanadium-titanium-iron occurrences, 13 PGE occurrences and 38 nickel-copper occurrences

  15. Analysis of some fuel characteristics deviations and their influence over WWER-440 fuel cycle design

    International Nuclear Information System (INIS)

    Stoyanova, I.; Kamenov, K.

    2001-01-01

    The aim of this study is to estimate the influence of some deviations in WWER-440 fuel assemblies (FA) characteristics upon fuel core design. A large number of different fresh fuel assemblies with enrichment of 3.5 t % are examined related to the enrichment, mass of initial metal Uranium and assembly shroud thickness. Infinite multiplication factor (Kinf) in fuel assembly has been calculated by HELIOS spectral code for basic assembly and for different FA with deviation of a single parameter. The effects from single parameter deviation (enrichment) and from two parameter deviations (enrichment and wall thickness) on the neutron-physics characteristics of the core are estimated for different fuel assemblies. Relatively week burnup dependence on Kinf is observed as result of deviation in the enrichment of the fuel and in the wall thickness of the assembly. An assessment of a FA single and two parameter deviations effects on design fuel cycle duration and relative power peaking factor is also considers in the paper. As a final conclusion can be settled that the maximum relative shortness of fuel cycle can be observed in the case of two FA parameters deviations

  16. Basic study on characteristics of some important equilibrium fuel cycles of PWR

    International Nuclear Information System (INIS)

    Waris, A.; Sekimoto, H.

    2001-01-01

    Equilibrium fuel cycle characteristics of a light water reactor (LWR) with enriched uranium supply were evaluated. In this study, five kinds of fuel cycles of 3000 MWt pressurized water reactor (PWR) were investigated, and a method to determine the uranium enrichment in order to achieve their criticality was presented. The results show that the enrichment decreases considerably with increasing number of confined heavy nuclides when U is discharged from the reactor. The required natural uranium was also evaluated for two different enrichment processes. The amount of required natural uranium also decreases as well. On the other hand, when U is totally confined, the enrichment becomes unacceptably high. Furthermore, Pu and minor actinides (MA) confining seem effective to incinerate the discharged radio-toxic wastes

  17. A study on the safety of spent fuel management. A scenario study on spent fuel management

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-03-01

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

  18. Nuclear fuel behavior activities at the OECD/NEA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The work programme regarding nuclear fuel behavior issues at OECD/NEA is carried out in two sections. The Nuclear Science and Data Bank Division deals with basic phenomena in fuel behavior under normal operating conditions, while the Safety Division concentrates upon regulation and safety issues in fuel behavior. A new task force addressing these latter issues has been set up and will produce a report providing recommendations in this field. The OECD Nuclear Energy Agency jointly with the International Atomic Energy Agency established an International Fuel Performance Experiments Database which is operated by the NEA Data Bank. (author). 1 tab.

  19. Nuclear fuel behavior activities at the OECD/NEA

    International Nuclear Information System (INIS)

    1997-01-01

    The work programme regarding nuclear fuel behavior issues at OECD/NEA is carried out in two sections. The Nuclear Science and Data Bank Division deals with basic phenomena in fuel behavior under normal operating conditions, while the Safety Division concentrates upon regulation and safety issues in fuel behavior. A new task force addressing these latter issues has been set up and will produce a report providing recommendations in this field. The OECD Nuclear Energy Agency jointly with the International Atomic Energy Agency established an International Fuel Performance Experiments Database which is operated by the NEA Data Bank. (author). 1 tab

  20. Nagra technical report 14-02, geological basics - Dossier V - Hydro-geological conditions

    International Nuclear Information System (INIS)

    Traber, D.; Gautschi, A.; Marschall, P.; Becker, J.; Waber, N.

    2014-01-01

    This dossier is the fifth of a series of eight reports concerning the safety and technical aspects of locations for the disposal of radioactive wastes in Switzerland. Dossier V looks at hydro-geological considerations in North-Western Switzerland. Rock layers in the region and their hydrological properties are examined. Ground-water and deeper lying aquifers in the various rock formations are discussed. The specific hydrology in the proposed areas for nuclear waste depositories is looked at, including infiltration and exfiltration zones and gradients

  1. Main conditions and effectiveness of gas fuel use for powering of dual fuel IC self-ignition engine

    Directory of Open Access Journals (Sweden)

    Stefan POSTRZEDNIK

    2015-09-01

    Full Text Available Internal combustion engines are fuelled mostly with liquid fuels (gasoline, diesel. Nowadays the gaseous fuels are applied as driving fuel of combustion engines. In case of spark ignition engines the liquid fuel (petrol can be totally replaced by the gas fuels. This possibility in case of compression engines is essentially restricted through the higher self-ignition temperatures of the combustible gases in comparison to classical diesel oil. Solution if this problem can be achieved by using of the dual fuel system, where for ignition of the prepared fuel gas - air mixture a specified amount of the liquid fuel (diesel oil should be additionally injected into the combustion chamber. For assurance that the combustion process proceeds without mistakes and completely, some basic conditions should be satisfied. In the frame of this work, three main aspects of this problem are taken into account: a. filling efficiency of the engine, b. stoichiometry of the combustion, c. performance of mechanical parameters (torque, power. A complex analysis of these conditions has been done and some achieved important results are presented in the paper.

  2. Conceptual design and cost estimation of dry cask storage facility for spent fuel

    International Nuclear Information System (INIS)

    Maki, Yasuro; Hironaga, Michihiko; Kitano, Koichi; Shidahara, Isao; Shiomi, Satoshi; Ohnuma, Hiroshi; Saegusa, Toshiari

    1985-01-01

    In order to propose an optimum storage method of spent fuel, studies on the technical and economical evaluation of various storage methods have been carried out. This report is one of the results of the study and deals with storage facility of dry cask storage. The basic condition of this work conforms to ''Basic Condition for Spent Fuel Storage'' prepared by Project Group of Spent Fuel Dry Storage at July 1984. Concerning the structural system of cask storage facilities, trench structure system and concrete silo system are selected for storage at reactor (AR), and a reinforced concrete structure of simple design and a structure with membrance roof are selected for away from reactor (AFR) storage. The basic thinking of this selection are (1) cask is put charge of safety against to radioactivity and (2) storage facility is simplified. Conceptual designs are made for the selected storage facilities according to the basic condition. Attached facilities of storage yard structure (these are cask handling facility, cask supervising facility, cask maintenance facility, radioactivity control facility, damaged fuel inspection and repack facility, waste management facility) are also designed. Cost estimation of cask storage facility are made on the basis of the conceptual design. (author)

  3. Outline of results of safety research (in nuclear fuel cycle field in fiscal year 1996)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The safety research in Power Reactor and Nuclear Fuel Development Corporation in fiscal year 1996 has been carried out based on the basic plan of safety research (from fiscal year 1996 to 2000) which was decided in March, 1996. In this report, on nuclear fuel cycle field, namely all the subjects in the fields of nuclear fuel facilities, environmental radioactivity and waste disposal, and the subjects related to nuclear fuel facilities among the fields of aseismatic and probabilistic safety assessments, the results of research in fiscal year 1996, the first year of the 5-year project, are summarized together with the outline of the basic plan of safety research. The basic policy, objective and system for promotion of the safety research are described. The objectives of the safety research are the advancement of safety technology, the safety of facilities, stable operation techniques, the safety design and the evaluation techniques of next generation facilities, and the support of transferring nuclear fuel cycle to private businesses. The objects of the research are uranium enrichment, fuel fabrication and reprocessing, and waste treatment and storage. 52 investigation papers of the results of the safety research in nuclear fuel cycle field in fiscal year 1996 are collected in this report. (K.I.)

  4. Basic properties of a zirconia based fuel material for LWRs

    International Nuclear Information System (INIS)

    Degueldre, C.; Paratte, J.M.

    1997-01-01

    The properties of zirconia cubic solid solutions doped with yttria, erbia and ceria or thoria are investigated with emphasis on the potential use of this material as inert matrix fuel for plutonium incineration in a light water reactor (LWR). The material is selected on the basis of its neutronic properties. Zr and Y are not neutron absorbers. Among the rare earth elements, Er was identified as a suitable burnable poison. The high density cubic solid solution is stable for a rather large range of compositions and from room temperature up to about 3000 K. Samples irradiated under low and high energy Xe ion irradiation up to a fluence of 1.8.10 16 Xe.cm -2 were investigated by transmission electron microscopy. Low energy (60 keV) Xe ions did not produce amorphization. From the observed bubble formation, swelling values during irradiation at room temperature or at high temperature (925 K) were estimated to be 0.1-0.72% by volume. Furthermore, no amorphization was obtained by Xe irradiation under extreme conditions such as high energy (1.5 MeV) Xe ion irradiation and low temperature (20 K). This confirms the robustness of this material and argues in favour of the selection of a zirconia based material as an advanced nuclear fuel for plutonium incineration. (author) 5 figs., 1 tab., 17 refs

  5. Learning FuelPHP for effective PHP development

    CERN Document Server

    Tweedie, Ross

    2013-01-01

    The book follows a standard tutorial approach, which will enable readers to use the FuelPHP framework efficiently while developing PHP applications.If you are a PHP developer who is looking to learn more about using the FuelPHP framework for effective PHP development, this book is ideal for you. If you are interested in this book, you should already have a basic understanding of general PHP development.

  6. Expansion of capacity of spent fuel pools and associated problems

    International Nuclear Information System (INIS)

    Francisco, J.L. De; Lopez-Cotarelo, J.; Ramos, J.M.

    1978-01-01

    Expanding the spent fuel storage pool capacity is a good solution for utilities facing the current shortage in fuel reprocessing capacity. The problems more likely to be found when expanding a spent fuel storage facility by using high density storage racks are reviewed. Basically three types of problems arise: 1) Problems related with the characteristics of the new facility. 2) Problems related with the works of expansion. 3) Problems related with the long term storage of large quantities of spent fuel. (author)

  7. Combustion from basics to applications

    CERN Document Server

    Lackner, Maximilian; Winter, Franz

    2013-01-01

    Combustion, the process of burning, is defined as a chemical reaction between a combustible reactant (the fuel) and an oxidizing agent (such as air) in order to produce heat and in most cases light while new chemical species (e.g., flue gas components) are formed. This book covers a gap on the market by providing a concise introduction to combustion. Most of the other books currently available are targeted towards the experienced users and contain too many details and/or contain knowledge at a fairly high level. This book provides a brief and clear overview of the combustion basics, suitable f

  8. New techniques for the characterization of refuse-derived fuels and solid recovered fuels.

    Science.gov (United States)

    Rotter, Vera Susanne; Lehmann, Annekatrin; Marzi, Thomas; Möhle, Edda; Schingnitz, Daniel; Hoffmann, Gaston

    2011-02-01

    Solid recovered fuel (SRF) today refers to a waste-derived fuel meeting defined quality specifications, in terms of both origin (produced from non-hazardous waste) and levels of certain fuel properties. Refuse-derived fuel (RDF) nowadays is more used for unspecified waste after a basic processing to increase the calorific value and therefore this term usually refers to the segregated, high calorific fraction of municipal solid waste (MSW), commercial or industrial wastes. In comparison with conventional fuels, both types of secondary fuel show waste of inherently varying quality and an increased level of waste-specific contaminants.The transition from RDF to SRF in the emerging national and European market requires a quality assurance system with defined quality parameters and analytical methods to ensure reliable fuel characterization. However, due to the quality requirements for RDF and SRF, the current standardized analysis methods often do not meet these practical demands. Fast test methods, which minimize personnel, financial and time efforts and which are applicable for producers as well as users can be an important supporting tool for RDF- and SRF-characterization. Currently, a fast test system based on incineration and correlation analyses which enable the determination of relevant fuel parameters is under development. Fast test methods are not aimed at replacing current standardized test methods, but have to be considered as practical supporting tools for the characterization of RDF and SRF.

  9. CFD Simulation of Heat and Fluid Flow for Spent Fuel in a Dry Storage

    International Nuclear Information System (INIS)

    In, Wangkee; Kwack, Youngkyun; Kook, Donghak; Koo, Yanghyun

    2014-01-01

    A dry storage system is used for the interim storage of spent fuel prior to permanent depository and/or recycling. The spent fuel is initially stored in a water pool for more than 5 years at least after dispatch from the reactor core and is transported to dry storage. The dry cask contains a multiple number of spent fuel assemblies, which are cooled down in the spent fuel pool. The dry cask is usually filled up with helium gas for increasing the heat transfer to the environment outside the cask. The dry storage system has been used for more than a decade in United States of America (USA) and the European Union (EU). Korea is also developing a dry storage system since its spent fuel pool is anticipated to be full within 10 years. The spent fuel will be stored in a dry cask for more than 40 years. The integrity and safety of spent fuel are important for long-term dry storage. The long-term storage will experience the degradation of spent fuel such as the embrittlement of fuel cladding, thermal creep and hydride reorientation. High burn-up fuel may expedite the material degradation. It is known that the cladding temperature has a strong influence on the material degradation. Hence, it is necessary to accurately predict the local distribution of the cladding temperature using the Computational Fluid Dynamics (CFD) approach. The objective of this study is to apply the CFD method for predicting the three-dimensional distribution of fuel temperature in a dry cask. This CFD study simulated the dry cask for containing the 21 fuel assemblies under development in Korea. This paper presents the fluid velocity and temperature distribution as well as the fuel temperature. A two-step CFD approach was applied to simulate the heat and fluid flow in a dry storage of 21 spent fuel assemblies. The first CFD analysis predicted the helium flow and temperature in a dry cask by a assuming porous body of the spent fuel. The second CFD analysis was to simulate a spent fuel assembly in the

  10. Nuclear fuel utilization in Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Boyadzhiev, Z; Kharalampieva, Ts; Pejchinov, Ts [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1994-12-31

    An assessment of fuel utilization in Kozloduy NPP units 1-6 is made on the basis of operational data obtained for a total of 62 fuel cycles. Basic characteristics of core loading and operation conditions are given. SPPS-1 and BIPR-7 codes are used to calculate assembly-wise power distributions for different full power days of a given cycle and unit. The data are compared with the measured values of these quantities. The analysis performed shows that the core loading option chosen has led to efficient fuel utilization without violation of the nuclear safety criteria. For WWER-440 (Units 1 - 4) this is expressed in effective reduction of the reactor vessel irradiation, maintaining the design duration of the fuel cycles at a reduced number of assemblies by a factor 5 - 5-10%, utilizing fuel with higher enrichment and implementing the 4-year fuel cycle. For WWER-1000 the improvements lead to: adoption of the 3-year fuel cycle utilizing fuel with 4.4% initial enrichment, implementation of improved fuel with a new type of absorbers and more effective low-leakage core loading patterns. 10 tabs., 6 figs., 7 refs.

  11. Fluid flow test for KMRR fuel assemblies

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Yang, Sun Kyu; Chung, Chang Hwan; Chun, See Young; Song, Chul Hha; Jun, Hyung Gil; Chung, Heung Joon; Won, Soon Yeun; Cho, Young Rho; Kim, Bok Deuk

    1991-01-01

    Hydraulic and velocity measurment tests were carried out for the KMRR fuel assembly. Two types of the KMRR fuel assembly are consist of longitudinally finned rods. Experimental data of the pressure drops and friction factors for the KMRR fuel assemlby were produced. The measurement technique for the turbulent flow structure in subchannels using the LDV was obtained. The measurement of the experimental constant of the thermal hydraulic analysis code was investigated. The results in this study are used as the basic data for the development of an analysis code. The measurement technique acquired in this study can be applied to the KMRR thermal hydraulic commissioning test and development of the domestic KMRR fuel fabrication. (Author)

  12. Development of nuclear fuel cycle technology

    International Nuclear Information System (INIS)

    Kawahara, Akira; Sugimoto, Yoshikazu; Shibata, Satoshi; Ikeda, Takashi; Suzuki, Kazumichi; Miki, Atsushi.

    1990-01-01

    In order to establish the stable supply of nuclear fuel as an important energy source, Hitachi ltd. has advanced the technical development aiming at the heightening of reliability, the increase of capacity, upgrading and the heightening of performance of the facilities related to nuclear fuel cycle. As for fuel reprocessing, Japan Nuclear Fuel Service Ltd. is promoting the construction of a commercial fuel reprocessing plant which is the first in Japan. The verification of the process performance, the ensuring of high reliability accompanying large capacity and the technical development for recovering effective resources from spent fuel are advanced. Moreover, as for uranium enrichment, Laser Enrichment Technology Research Association was founded mainly by electric power companies, and the development of the next generation enrichment technology using laser is promoted. The development of spent fuel reprocessing technology, the development of the basic technology of atomic process laser enrichment and so on are reported. In addition to the above technologies recently developed by Hitachi Ltd., the technology of reducing harm and solidification of radioactive wastes, the molecular process laser enrichment and others are developed. (K.I.)

  13. Agricultural transportation fuels

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    The recommendations on the title subject are focused on the question whether advantages and disadvantages of agricultural fuels compared to fossil fuels justify the Dutch policy promotion of the use of agricultural products as basic materials for agricultural fuels. Attention is paid to energetic, environmental and economical aspects of both fuel types. Four options to apply agricultural transportation fuels are discussed: (1) 10% bio-ethanol in euro-unleaded gasoline for engines of passenger cars, equipped with a three-way catalyst; (2) the substitution of 15% methyl tertiair butyl ether (MTBE) by ethyl tertiair butyl ether (ETBE) as a substituent for lead in unleaded super plus gasoline (Sp 98) for engines of passenger cars, equipped with a three-way catalyst; (3) 50% KME (rapeseed oil ester) in low-sulfur diesel (0.05%S D) for engines of vans without a catalyst; and (4) the substitution of 0.05% S D by bio-ethanol or KME for buses with fuel-adjusted engines, equipped with a catalyst. Also the substitution by liquefied petroleum gas (LPG), compressed natural gas (CNG) or E 95 was investigated in option four. Each of the options investigated can contribute to a reduction of the use of fossil energy and the environmental effects of the use of fossil fuels, although some environmental effects from agricultural fuels must be taken into consideration. It is recommended to seriously pay attention to the promotion of agricultural fuels, not only in the Netherlands, but also in an international context. Policy instruments to be used in the stimulation of the use of such fuels are the existing European Community subsidies on fallow lands, exemption of the European Community energy levy, and the use of tax differentiation. Large-scale demonstration projects must be started to quantify hazardous emissions and to solve still existing technical problems. 8 figs., 3 tabs., refs., 4 appendices

  14. Corrosion Behaviour of Mg Alloys in Various Basic Media: Application of Waste Encapsulation of Fuel Decanning from UNGG Nuclear Reactor

    Science.gov (United States)

    Lambertin, David; Frizon, Fabien; Blachere, Adrien; Bart, Florence

    The dismantling of UNGG nuclear reactor generates a large volume of fuel decanning. These materials are based on Mg-Zr alloy. The dismantling strategy could be to encapsulate these wastes into an ordinary Portland cement (OPC) or geopolymer (aluminosilicate material) in a form suitable for storage. Studies have been performed on Mg or Mg-Al alloy in basic media but no data are available on Mg-Zr behaviour. The influence of representative pore solution of both OPC and geopolymer with Mg-Zr alloy has been studied on corrosion behaviour. Electrochemical methods have been used to determine the corrosion densities at room temperature. Results show that the corrosion densities of Mg-Zr alloy in OPC solution is one order of magnitude more important than in a geopolymer solution environment and the effect of an inhibiting agent has been undertaken with Mg-Zr alloy. Evaluation of corrosion hydrogen production during the encapsulation of Mg-Zr alloy in both OPC and geopolymer has also been done.

  15. Domestic fuel question and the charcoal solution

    Energy Technology Data Exchange (ETDEWEB)

    Krishna Rao, E G

    1981-06-01

    Domestic fuel for cooking forms one of the basic needs of human society. In India, the pressure of this need has exceeded the regeneration potential of the growing forests which supply a large proportion of this basic need. The pressure can be greatly relieved by converting wood to charcoal before it reaches the consumer. The present paper examines this aspect and reviews the modern methods of charcoal production on fuelwood resources. Besides being a choice domestic fuel, charcoal is a valuable raw material in various industries. Charcoal making industry can be established as a rural based industry (as part of community forestry projects) and would generate much needed cash income at grassroot level. The strategy would be important in dealing with the problem of chronic poverty at this level. (Refs. 5).

  16. Integral benchmarks with reference to thorium fuel cycle

    International Nuclear Information System (INIS)

    Ganesan, S.

    2003-01-01

    This is a power point presentation about the Indian participation in the CRP 'Evaluated Data for the Thorium-Uranium fuel cycle'. The plans and scope of the Indian participation are to provide selected integral experimental benchmarks for nuclear data validation, including Indian Thorium burn up benchmarks, post-irradiation examination studies, comparison of basic evaluated data files and analysis of selected benchmarks for Th-U fuel cycle

  17. Design premises for canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    Werme, L.

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel

  18. The atomic energy basic law

    International Nuclear Information System (INIS)

    1977-01-01

    The law establishes clearly the principles that Japan makes R and D, and utilizations of atomic energy only for the peaceful purposes. All the other laws and regulations concerning atomic energy are based on the law. The first chapter lays down the above mentioned objective of the law, and gives definitions of basic concepts and terms, such as atomic energy, nuclear fuel material, nuclear source material, nuclear reactor and radiation. The second chapter provides for the establishment of Atomic Energy Commission which conducts plannings and investigations, and also makes decisions concerning R and D, and utilizations of atomic energy. The third chapter stipulates for establishment of two government organizations which perform R and D of atomic energy developments including experiments and demonstrations of new types of reactors, namely, Atomic Energy Research Institute and Power Reactor and Nuclear Fuel Development Corporation. Chapters from 4th through 8th provide for the regulations on development and acquisition of the minerals containing nuclear source materials, controls on nuclear fuel materials and nuclear reactors, administrations of the patents and inventions concerning atomic energy, and also prevention of injuries due to radiations. The last 9th chapter requires the government and its appointee to compensate the interested third party for damages in relation to the exploitation of nuclear source materials. (Matsushima, A.)

  19. A study on the safety of spent fuel management. Radioactive source term modelling

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Kwan Sik; Lee, Hoo Keun; Park, Keun Il; Hwoang, Jung Ki; Chung, Choong Hwan [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

    1992-02-01

    The types and probabilities of events which may occur during the process of reception, transfer and storage of spent fuels in an away-from-reactor (AFR) spent fuel storage facility were analyzed in order to calculate the amount of radioactive material released to operation area and atmosphere, and the basic model for predicting the radioactive source-term under normal and abnormal operations were developed. Also, oxidation and dissolution of U0{sub 2} pellet was investigated to estimate the amount of radioactive materials released from spent fuel and the release characteristics of radionuclides from defected spent fuel rods was analyzed. Basic information using FIRAC code to analyze the ventilation system during fire accident was prepared and FIRIN was detached from FIRAC modified to simulate the compartment fire by personal computer. (Author).

  20. Development of Advanced High Uranium Density Fuels for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

    2016-02-15

    This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

  1. PEM fuel cell modeling and simulation using Matlab

    CERN Document Server

    Spiegel, Colleen

    2011-01-01

    Although, the basic concept of a fuel cell is quite simple, creating new designs and optimizing their performance takes serious work and a mastery of several technical areas. PEM Fuel Cell Modeling and Simulation Using Matlab, provides design engineers and researchers with a valuable tool for understanding and overcoming barriers to designing and building the next generation of PEM Fuel Cells. With this book, engineers can test components and verify designs in the development phase, saving both time and money.Easy to read and understand, this book provides design and modelling tips for

  2. 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

  3. Some possibilities for improvement of fuel utilization in nuclear power plants

    International Nuclear Information System (INIS)

    Kocic, A.; Marinkovic, N.

    1983-01-01

    Methods for improving the nuclear fuel utilization with the emphasis on LWRs are being dealt with in this paper. Some basic results concerning tubular fuel pellets of the Krsko nuclear power plants are presented, showing promising possibilities for uranium saving from the neutronics point of view. (author)

  4. Fuel supply security

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki

    1987-01-01

    Stable fuel supply is a prerequisite for any nuclear power program including ISER-PIUS. It encompasses procurement of uranium ore, enriched uranium and fuel elements. Uranium is different from oil in that it can be stockpiled for more than a decade besides the fact that the core residence time is as long as six years, for example in the case of ISER-PIUS. These basic fuel characteristics are favoring nuclear fuel over others in terms of supply security. The central concern will be a gradual increase in prices of uranium and enrichment. Under the present glut situation with the worldwide prevalence of LWRs, fuel supply security seems ensured for the time being till the middle of 21st century. It is estimated that by the turn of the century, the free world will have roughly 450 GWe capacity of nuclear power. If 10 % is supplied for ISER-PIUS, more than 200 modules of 200 MWe ISER-PIUS may be deployed all over the world probably starting around 2000. As part of the fuel supply security consideration, heavy water reactor (HWR) may seem interesting to such a country as Indonesia where there is uranium resources but no enrichment capability. But it needs heavy water instead and the operation is not so easy as of LWR, because of the positive void coefficient as was seen at the Chernobyl-4. Safeguarding of the fuel is also difficult, because it lends itself to on line refueling. The current and future situation of the fuel supply security for LWR seem well founded and established long into the future. (Nogami, K.)

  5. Basic mechanisms of photosynthesis and applications to improved production and conversion of biomass to fuels and chemical products

    Energy Technology Data Exchange (ETDEWEB)

    El-Sayed, M. [Georgia Institute of Tech., Atlanta, GA (United States); Greenbaum, E. [Oak Ridge National Laboratory, TN (United States); Wasielewski, M. [Argonne National Lab., IL (United States)

    1996-09-01

    Natural photosynthesis, the result of 3.5 billion years of evolutionary experimentation, is the best proven, functional solar energy conversion technology. It is responsible for filling the vast majority of humanity`s energy, nutritional, and materials needs. Understanding the basic physical chemical principles underlying photosynthesis as a working model system is vital to further exploitation of this natural technology. These principles can be used to improve or modify natural photosynthesis so that it is more efficient or so that it can produce unusual products such as hydrogen, methane, methanol, ethanol, diesel fuel substitutes, biodegradable materials, or other high value chemical products. Principles garnered from the natural process can also be used to design artificial photosynthetic devices that employ analogs of natural antenna and reaction center function, self-assembly and repair concepts, photoinduced charge transfer processes, photoprotection, and dark reactions that facilitate catalytic action to convert light into, useful chemical or electrical energy. The present broad understanding of many structural and functional aspects of photosynthesis has resulted from rapid recent research progress. X-ray structures of several key photosynthetic reaction centers and antenna systems are available, and the overall principles controlling photoinduced energy and electron transfer are being established.

  6. Model of cooling nuclear fuel rod in the nuclear reactor

    International Nuclear Information System (INIS)

    Lavicka, David; Polansky, Jiri

    2010-01-01

    The following topics are described: Some basic requirements for nuclear fuel rods; The VVER 1000 fuel rod; Classification of the two-phase flow in the vertical tube; Type of heat transfer crisis in the vertical tube; Experimental apparatus; Model of the nuclear fuel rod and spacers; Potential of the experimental apparatus (velocity profile measurement via PIV; thermal flow field measurement by the PLIF method; cooling graph in dependence on the fuel rod temperature; comparison of the hydrodynamic properties with respect to the design features of the spacers). (P.A.)

  7. Spent fuel storage for ISER plant

    International Nuclear Information System (INIS)

    Nakajima, Takasuke; Kimura, Yuzi

    1987-01-01

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

  8. Application of game theory in decision making strategy: Does gas fuel industry need to kill oil based fuel industry?

    Science.gov (United States)

    Azmi, Abdul Luky Shofi'ul; Prabandari, Dyah Lusiana; Hakim, Muhammad Lintang Islami

    2017-03-01

    Even though conversion of oil based fuel (Bahan Bakar Minyak) into gas fuel (Bahan Bakar Gas) for transportation (both land and sea) is one of the priority programs of the government of Indonesia, rules that have been established merely basic rules of gas fuel usage license for transportation, without discussing position of gas fuel related to oil based fuel in detail. This paper focus on possible strategic behavior of the key players in the oil-gas fuel conversion game, who will be impacted by the position of gas fuel as complement or substitution of oil based fuel. These players include industry of oil based fuel, industry of gas fuel, and the government. Modeling is made based on two different conditions: government plays a passive role and government plays an active role in legislating additional rules that may benefit industry of gas fuel. Results obtained under a passive government is that industry of oil based fuel need to accommodate the presence of industry of gas fuel, and industry of gas fuel does not kill/ eliminate the oil based fuel, or gas fuel serves as a complement. While in an active government, the industry of oil based fuel need to increase its negotiation spending in the first phase so that the additional rule that benefitting industry of gas fuel would not be legislated, while industry of gas fuel chooses to indifferent; however, in the last stage, gas fuel turned to be competitive or choose its role to be substitution.

  9. National Policy on Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Soedyartomo, S.

    1996-01-01

    National policy on nuclear fuel cycle is aimed at attaining the expected condition, i.e. being able to support optimality the national energy policy and other related Government policies taking into account current domestic nuclear fuel cycle condition and the trend of international nuclear fuel cycle development, the national strength, weakness, thread and opportunity in the field of energy. This policy has to be followed by the strategy to accomplish covering the optimization of domestic efforts, cooperation with other countries, and or purchasing licences. These policy and strategy have to be broken down into various nuclear fuel cycle programmes covering basically assesment of the whole cycle, performing research and development of the whole cycle without enrichment and reprocessing being able for weapon, as well as programmes for industrialization of the fuel cycle stepwisery commencing with the middle part of the cycle and ending with the edge of the back-end of the cycle

  10. The state-of-the-art and problems of fuel element structural analysis

    International Nuclear Information System (INIS)

    Lassmann, K.

    1980-02-01

    This study of fuel element structural analysis is arranged in two parts: In the first, self-contained, part the general basic principles of deterministic computer programs for structural analysis of fuel elements are reviewed critically and an approach is shown which can be used to expand the system with respect to statistical investigations. The second part contains technical details summarized in 11 publications, all of which appeared in periodicals with reviewer teams. The major aspects of this study are thought to be the following ones: Contributions to the 'philosophy' of fuel element structural analysis. Critical analysis of the basic structure of computer programs. Critical analysis of the mechanical concept of integral fuel rod computer programs. Establishment of a comprehensive computer program system (URANUS). Expansion from purely deterministic information by statistical analyses. Methodological and computer program developments for the analysis of fast accidents. (orig.) 891 HP/orig. 892 MKO [de

  11. Agro-fuels, a cartography of stakes

    International Nuclear Information System (INIS)

    2008-09-01

    This document proposes a dashboard of the main issues regarding agro-fuels. Nine sheets propose basic information and data on these issues: 1- agro-fuel production and consumption in the world (ethanol, vegetable oils, perspective for demand in the transport sector), 2- energy efficiency and greenhouse gas emissions (energy assessments and greenhouse effect of agro-fuels, discrepancies of results between first-generation European agro-fuels, case of agro-fuels produced in Southern countries), 3- needed surfaces in Europe (land use and cultivable areas for agro-fuel production in Europe and in France, competition between food and energy crops), 4- deforestation in the South (relationship between agriculture, deforestation and agro-fuels, between deforestation and greenhouse gas emissions), 5- impacts on biodiversity (use of pesticides and fertilizers, large scale cultivations and single-crop farming, cultivation of fallow land and permanent meadows, deforestation in the South, relationship between agro-fuels and GMOs), 6- impacts on water, soil and air (water quality and availability, soil erosion, compaction and fertility loss, air quality), 7- food-related and social stakes (issue of food security, social impacts of agro-fuel production with pressure on family agriculture and issues of land property), 8- public supports and economic efficiency (public promotion of agro-fuels, agro-fuel and oil prices, assessment of the 'avoided' CO 2 ton), and 9- perspectives for second-generation agro-fuels (definitions and processes, benefits with respect to first-generation fuels, possible impacts on the environment, barriers to their development)

  12. Design premises for canister for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Werme, L

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel 43 refs, 4 figs, 6 tabs

  13. Analysis of radwaste material management options for experimental DUPIC fuel fabrication process

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Park, J. J.; Yang, M. S.; Kim, K. H.; Shin, J. M.; Lee, H. S.; Ko, W. I.; Lee, J. W.; Yim, S. P.; Hong, D. H.; Lee, J. Y.; Baik, S. Y.; Song, W. S.; Yoo, B. O.; Lee, E. P.; Kang, I. S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    This report is desirable to review management options in advance for radioactive waste generated from manufacturing experiment of DUPIC nuclear fuel as well as residual nuclear material and dismantled equipment. This report was written for helping researchers working in related facilities to DUPIC project understanding management of DUPIC radioactive waste as well as fellows in DUPIC project. Also, it will be used as basic material to prove transparency and safeguardability of DUPIC fuel cycle. In order to meet these purposes, this report includes basic experiment plan for manufacturing DUPIC nuclear fuel, outlines for DUPIC manufacturing facility and equipment, arising source and estimated amount of radioactive waste, waste classification and packing, transport cask, transport procedures. 15 refs., 31 figs., 11 tabs. (Author)

  14. Corrosion surveillance programme for Latin American research reactor Al-clad spent fuel in water

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Haddad, R.; Ritchie, I.

    2002-01-01

    The objectives of the IAEA sponsored Regional Technical Co-operation Project for Latin America (Argentina, Brazil, Chile, Mexico, and Peru) are to provide the basic conditions to define a regional strategy for managing spent fuel and to provide solutions, taking into consideration the economic and technological realities of the countries involved. In particular, to determine the basic conditions for managing research reactor spent fuel during operation and interim storage as well as final disposal, and to establish forms of regional cooperation in the four main areas: spent fuel characterization, safety, regulation and public communication. This paper reports the corrosion surveillance activities of the Regional Project and these are based on the IAEA sponsored co-ordinated research project (CRP) on 'Corrosion of research reactor Al-clad spent fuel in water'. The overall test consists of exposing corrosion coupon racks at different spent fuel basins followed by evaluation. (author)

  15. Fuel processing for PEM fuel cells: transport and kinetic issues of system design

    Science.gov (United States)

    Zalc, J. M.; Löffler, D. G.

    In light of the distribution and storage issues associated with hydrogen, efficient on-board fuel processing will be a significant factor in the implementation of PEM fuel cells for automotive applications. Here, we apply basic chemical engineering principles to gain insight into the factors that limit performance in each component of a fuel processor. A system consisting of a plate reactor steam reformer, water-gas shift unit, and preferential oxidation reactor is used as a case study. It is found that for a steam reformer based on catalyst-coated foils, mass transfer from the bulk gas to the catalyst surface is the limiting process. The water-gas shift reactor is expected to be the largest component of the fuel processor and is limited by intrinsic catalyst activity, while a successful preferential oxidation unit depends on strict temperature control in order to minimize parasitic hydrogen oxidation. This stepwise approach of sequentially eliminating rate-limiting processes can be used to identify possible means of performance enhancement in a broad range of applications.

  16. Fight against fuel poverty. Levers, stakes and expectations of the fight against fuel poverty in housing

    International Nuclear Information System (INIS)

    Payen, Luc; Pamart, Isabelle; Lacroix, Olivier

    2013-10-01

    'Is in fuel poverty a person who feels in his particular housing difficulties have the necessary energy supply to the satisfaction of basic needs due to the inadequacy of resources or its habitat conditions'. The rising cost of energy commodities in the late 2000's, added to the poor thermal quality an important part of French homes, has led to the emergence of fuel poverty in the public debate. Legislative recognition of these situations with the law 'Grenelle II' (from which is extracted the definition above) marked a decisive step in the fight against this complex problem. Affecting nearly 5 million households in France, fuel poverty is a major challenge for societies wishing to successfully achieve their energy transition. In this new publication, ENEA reports on the main levers of the fight against fuel poverty, the obstacles encountered and the needs for new solutions

  17. The neutronic and fuel cycle performance of interchangeable 3500 MWth metal and oxide fueled LMRs

    International Nuclear Information System (INIS)

    Fujita, E.K.; Wade, D.C.

    1990-01-01

    This study summarizes the neutronic and fuel cycle analysis performed at Argonne National Laboratory for an oxide and a metal fueled 3500 MWth LMR. These reactor designs formed the basis for a joint US/European study of LMR ATWS events. The oxide and metal core designs were developed to meet reactor performance specifications that are constrained by requirements for core loading interchangeability and for a small burnup reactivity swing. Differences in the computed performance parameters of the oxide and metal cores, arising from basic differences in their neutronic characteristics, are identified and discussed. It is shown that metal and oxide cores designed to the same ground rules exhibit many similar performance characteristics; however, they differ substantially in reactivity coefficients, control strategies, and fuel cycle options. 12 refs., 2 figs., 12 tabs

  18. PEM fuel cells thermal and water management fundamentals

    CERN Document Server

    Wang, Yun; Cho, Sung Chan

    2014-01-01

    Polymer Electrolyte Membrane (PEM) fuel cells convert chemical energy in hydrogen into electrical energy with water as the only by-product. Thus, PEM fuel cells hold great promise to reduce both pollutant emissions and dependency on fossil fuels, especially for transportation-passenger cars, utility vehicles, and buses-and small-scale stationary and portable power generators. But one of the greatest challenges to realizing the high efficiency and zero emissions potential of PEM fuel cells technology is heat and water management. This book provides an introduction to the essential concepts for effective thermal and water management in PEM fuel cells and an assessment on the current status of fundamental research in this field. The book offers you: An overview of current energy and environmental challenges and their imperatives for the development of renewable energy resources, including discussion of the role of PEM fuel cells in addressing these issues; Reviews of basic principles pertaining to PEM fuel cel...

  19. Thermochemical data and its use in modeling chemical behavior in mixed-oxide fuel pins

    International Nuclear Information System (INIS)

    Gibby, R.L.; Woodley, R.E.; Adamson, M.G.; Johnson, C.E.

    1979-01-01

    The status of US activities to obtain fuel chemistry data is reviewed. Analytical expressions addressing basic needs of all fuel chemistry models are presented. Fission product concentrations during irradiation, oxygen-to-metal (O/M) at beginning-of-life and at burnup, and the potential in fuel-cladding gap at burnup are described

  20. Practice and trends in nuclear fuel licensing in France (pressurized water reactor fuels)

    International Nuclear Information System (INIS)

    Roudier, S.; Badel, D.; Beraha, R.; Champ, M.; Tricot, N.; Tran Dai, P.

    1994-01-01

    The activities of governmental French authorities responsible for safety of nuclear installations are outlined. The main bodies involved in nuclear safety are: the CSSIN (High Council for Nuclear Safety and Information), CINB (Inter-ministerial Commission for Basic Nuclear Installations) and DSIN (Nuclear Installations Safety Directorate). A brief review of the main fuel licensing issues supported by DSIN is given, which includes: 1) formal regularity procedure ensuring the safety of nuclear installations and especially the pressurized water reactors; 2) guidelines for nuclear design and manufacturing requirements related to safety and 3) safety goals and associated limits. The fuel safety documents for reloading as well as the research and development programmes in the field of technical safety are also described. The ongoing experiments in CABRI reactor, aimed at determining the high burnup fuel behaviour under reactivity initiated accidents until 65 GW d/Mt U, are one of these programs

  1. Practice and trends in nuclear fuel licensing in France (pressurized water reactor fuels)

    Energy Technology Data Exchange (ETDEWEB)

    Roudier, S [Direction de la Surete des Installations Nucleaires, Fontenay-aux-Roses (France); Badel, D; Beraha, R [Direction Regionale de l` Industrie, de la Recherche et de l` Environnement Rhone-Alpes, Lyon (France); Champ, M; Tricot, N; Tran Dai, P [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire

    1994-12-31

    The activities of governmental French authorities responsible for safety of nuclear installations are outlined. The main bodies involved in nuclear safety are: the CSSIN (High Council for Nuclear Safety and Information), CINB (Inter-ministerial Commission for Basic Nuclear Installations) and DSIN (Nuclear Installations Safety Directorate). A brief review of the main fuel licensing issues supported by DSIN is given, which includes: (1) formal regularity procedure ensuring the safety of nuclear installations and especially the pressurized water reactors; (2) guidelines for nuclear design and manufacturing requirements related to safety and (3) safety goals and associated limits. The fuel safety documents for reloading as well as the research and development programmes in the field of technical safety are also described. The ongoing experiments in CABRI reactor, aimed at determining the high burnup fuel behaviour under reactivity initiated accidents until 65 GW d/Mt U, are one of these programs.

  2. History and current status of nuclear fuel reprocessing technology

    International Nuclear Information System (INIS)

    Funasaka, Hideyuki; Nagai, Toshihisa; Washiya, Tadahiro

    2008-01-01

    History and present state of fast breeder reactor was reviewed in series. As a history and current status of nuclear fuel reprocessing technology, this ninth lecture presented the progress of the FBR fuel reprocessing technology and advanced reprocessing processes. FBR fuel reprocessing technology had been developed to construct the reprocessing equipment test facilities (RETF) based on PUREX process technologies. With economics, reduction of environmental burdens and proliferation resistance taken into consideration, advanced aqueous method for nuclear fuel cycle activities has been promoted as the government's basic policy. Innovative technologies on mechanical disassembly, continuous rotary dissolver, crystallizer, solvent extraction and actinides recovery have been mainly studied. (T. Tanaka)

  3. Typical IAEA operations at a fuel fabrication plant

    International Nuclear Information System (INIS)

    Morsy, S.

    1984-01-01

    The IAEA operations performed at a typical Fuel Fabrication Plant are explained. To make the analysis less general the case of Low Enriched Uranium (LEU) Fuel Fabrication Plants is considered. Many of the conclusions drawn from this analysis could be extended to other types of fabrication plants. The safeguards objectives and goals at LEU Fuel Fabrication Plants are defined followed by a brief description of the fabrication process. The basic philosophy behind nuclear material stratification and the concept of Material Balance Areas (MBA's) and Key Measurement Points (KMP's) is explained. The Agency operations and verification methods used during physical inventory verifications are illustrated

  4. Probabilistic fuel rod analyses using the TRANSURANUS code

    Energy Technology Data Exchange (ETDEWEB)

    Lassmann, K; O` Carroll, C; Laar, J Van De [CEC Joint Research Centre, Karlsruhe (Germany)

    1997-08-01

    After more than 25 years of fuel rod modelling research, the basic concepts are well established and the limitations of the specific approaches are known. However, the widely used mechanistic approach leads in many cases to discrepancies between theoretical predictions and experimental evidence indicating that models are not exact and that some of the physical processes encountered are of stochastic nature. To better understand uncertainties and their consequences, the mechanistic approach must therefore be augmented by statistical analyses. In the present paper the basic probabilistic methods are briefly discussed. Two such probabilistic approaches are included in the fuel rod performance code TRANSURANUS: the Monte Carlo method and the Numerical Noise Analysis. These two techniques are compared and their capabilities are demonstrated. (author). 12 refs, 4 figs, 2 tabs.

  5. Surrounding rock stress analysis of underground high level waste repository

    International Nuclear Information System (INIS)

    Liu Wengang; Wang Ju; Wang Guangdi

    2006-01-01

    During decay of nuclear waste, enormous energy was released, which results in temperature change of surrounding rock of depository. Thermal stress was produced because thermal expansion of rock was controlled. Internal structure of surrounding rock was damaged and strength of rock was weakened. So, variation of stress was a dynamic process with the variation of temperature. BeiShan region of Gansu province was determined to be the depository field in the future, it is essential to make research on granite in this region. In the process of experiment, basic physical parameters of granite were analyzed preliminary with MTS. Long range temperature and stress filed was simulated considering the damage effect of surrounding rock, and rules of temperature and stress was achieved. (authors)

  6. Performance of refractory alloy-clad fuel pins

    International Nuclear Information System (INIS)

    Dutt, D.S.; Cox, C.M.; Millhollen, M.K.

    1984-12-01

    This paper discusses objectives and basic design of two fuel-cladding tests being conducted in support of SP-100 technology development. Two of the current space nuclear power concepts use conventional pin type designs, where a coolant removes the heat from the core and transports it to an out-of-core energy conversion system. An extensive irradiation testing program was conducted in the 1950's and 1960's to develop fuel pins for space nuclear reactors. The program emphasized refractory metal clad uranium nitride (UN), uranium carbide (UC), uranium oxide (UO 2 ), and metal matrix fuels (UCZr and BeO-UO 2 ). Based on this earlier work, studies presented here show that UN and UO 2 fuels in conjunction with several refractory metal cladding materials demonstrated high potential for meeting space reactor requirements and that UC could serve as an alternative but higher risk fuel

  7. Costing of spent nuclear fuel storage

    International Nuclear Information System (INIS)

    2009-01-01

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

  8. Modeling Thermal and Stress Behavior of the Fuel-clad Interface in Monolithic Fuel Mini-plates

    International Nuclear Information System (INIS)

    Miller, Gregory K.; Medvedev, Pavel G.; Burkes, Douglas E.; Wachs, Daniel M.

    2010-01-01

    As part of the Global Threat Reduction Initiative, a fuel development and qualification program is in process with the objective of qualifying very high density low enriched uranium fuel that will enable the conversion of high performance research reactors with operational requirements beyond those supported with currently available low enriched uranium fuels. The high density of the fuel is achieved by replacing the fuel meat with a single monolithic low enriched uranium-molybdenum fuel foil. Doing so creates differences in the mechanical and structural characteristics of the fuel plate because of the planar interface created by the fuel foil and cladding. Furthermore, the monolithic fuel meat will dominate the structural properties of the fuel plate rather than the aluminum matrix, which is characteristic of dispersion fuel types. Understanding the integrity and behavior of the fuel-clad interface during irradiation is of great importance for qualification of the new fuel, but can be somewhat challenging to determine with a single technique. Efforts aimed at addressing this problem are underway within the fuel development and qualification program, comprised of modeling, as-fabricated plate characterization, and post-irradiation examination. An initial finite element analysis model has been developed to investigate worst-case scenarios for the basic monolithic fuel plate structure, using typical mini-plate irradiation conditions in the Advanced Test Reactor. Initial analysis shows that the stress normal to the fuel-clad interface dominates during irradiation, and that the presence of small, rounded delaminations at the interface is not of great concern. However, larger and/or fuel-clad delaminations with sharp corners can create areas of concern, as maximum principal cladding stress, strain, displacement, and peak fuel temperature are all significantly increased. Furthermore, stresses resulting from temperature gradients that cause the plate to bow or buckle in

  9. About a fuel for burnup reactor of periodical pulsed nuclear pumped laser

    International Nuclear Information System (INIS)

    Volkov, A.I.; Lukin, A.V.; Magda, L.E.; Magda, E.P.; Pogrebov, I.S.; Putnikov, I.S.; Khmelnitsky, D.V.; Scherbakov, A.P.

    1998-01-01

    A physical scheme of burnup reactor for a Periodic Pulsed Nuclear Pumped Laser was supposed. Calculations of its neutron physical parameters were made. The general layout and construction of basic elements of the reactor are discussed. The requirements for the fuel and fuel elements are established. (author)

  10. MTR fuel plate qualification capabilities at SCK-CEN

    International Nuclear Information System (INIS)

    Koonen, E.; Jacquet, P.

    2002-01-01

    In order to enhance the capabilities of BR2 in the field of MTR fuel plate testing, a dedicated irradiation device has been designed. In its basic version this device allows the irradiation of 3 fuel plates. The central fuel plate may be replaced by a dummy plate or a plate carrying dosimeters. A first FUTURE device has been built. A benchmark irradiation has been executed with standard BR2 fuel plates in order to qualify this device. Detailed neutronic calculations were performed and the results compared to the results of the post-irradiation examinations of the plates. These comparisons demonstrate the capability to conduct a fuel plate irradiation program under requested and well-known irradiation conditions. Further improvements are presently being designed in order to extend the ranges of heat flux and surface temperature of the fuel plates that can be handled with the FUTURE device. (author)

  11. A novel H2S/H2O2 fuel cell operating at the room temperature

    Energy Technology Data Exchange (ETDEWEB)

    Sanli, Ayse Elif [Gazi University (Turkey)], email: aecsanli@gmail.com; Aytac, Aylin [Department of Chemistry, Faculty of Science, Gazi University, Teknikokullar (Turkey)], email: aytaca@gazi.edu.tr

    2011-07-01

    This study concerns the oxidation mechanism of hydrogen sulfide and a fuel cell; acidic peroxide is used as the oxidant and basic hydrogen sulfide is the fuel. A solid state H2S/H2O2 stable fuel cell was produced at room temperature. A cell potential of 0.85 V was reached; this is quite remarkable in comparison to the H2S/O2 fuel cell potential of 0.85 V obtained at 850-1000 degree celsius. The hydrogen sulfide goes through an oxidation reaction in the alkaline fuel cell (H2S/H2O2 fuel cell) which opens up the possibility of using the cheaper nickel as a catalyst. As a result, the fuel cell becomes a potentially low cost technology. A further benefit from using H2S as the alkaline liquid H2S/H2O2 fuel cell, is that sulfide ions are oxidized at the anode, releasing electrons. Sulfur produced reacts with the other sulfide ions and forms disulfide and polysulfide ions in basic electrolytes (such as Black Sea water).

  12. Materials in the environment of the fuel in dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Issard, H [TN International (Cogema Logistics) (France)

    2012-07-01

    Spent nuclear fuel has been stored safely in pools or dry systems in over 30 countries. The majority of IAEA Member States have not yet decided upon the ultimate disposition of their spent nuclear fuel: reprocessing or direct disposal. Interim storage is the current solution for these countries. For developing the technological knowledge data base, a continuation of the IAEA's spent fuel storage performance assessment was achieved. The objectives are: Investigate the dry storage systems and gather basic fuel behaviour assessment; Gather data on dry storage environment and cask materials; Evaluate long term behaviour of cask materials.

  13. The fuel reprocessing plant at Wackersdorf

    International Nuclear Information System (INIS)

    Held, M.

    1986-01-01

    For a more systematic discussion about the fuel reprocessing plant at Wackersdorf, the colloquium tried to cover the most important questions put forward in the controversies: economic efficiency and energy-political needs; safety and ecological repercussions; inner safety and consequences for basic rights and the regional economic structure; majority decisions and participation of the population of the region. Elements of evaluation are the conservation of resources, health, economic efficiency, and citizens' rights of liberty. The related basic ethical questions are considered. The 18 contributions are individually recorded in the data base. (DG) [de

  14. Health and safety of competing fuel options for fuel cells in the road transport sector

    Energy Technology Data Exchange (ETDEWEB)

    Green, E.; Short, S.; Stutt, E.; Wickramatillake, H.; Harrison, P.

    2000-07-01

    This report presents a critical analysis of the health and safety issues surrounding competing transport fuel options, including those for possible future fuel-cell powered vehicles. The fuels considered in this report are gasoline (unleaded and reformulated), diesel, hydrogen (H{sub 2}), methanol, natural gas and liquefied petroleum gas (LPG). The analysis presented here is based on available information in peer-reviewed, published papers and other sources such as government department or research laboratory reports and websites. An overall evaluation of the fuels in terms of their toxicity and health effects, environmental fate, and fire and explosion safety aspects is presented. The report is based on current knowledge and makes no assumptions as to how fuels may change in the future if they are to be used in fuel-cell vehicles. The report identifies the hazards of the fuels but does not estimate the risks likely to be associated with their eventual use in fuel-cell vehicles. The focus is on the fuels themselves and not their exhaust or reaction products. sNo assessment has been made of the environmental effects data for the fuels. Broad environmental considerations such as ozone forming potential and also global warming are not considered. Basic information on environmental fate is included to provide an understanding of migratory pathways, environmental compartmentalisation and potential routes of human exposure. Other factors such as economics, government incentives or disincentives and public attitudes may have a bearing on which of the fuels are considered most appropriate for future fuel-cell vehicles; these factors are not considered in any detail in this report. (Author)

  15. Germany's imports and exports of nuclear fuels in 1981

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    The statistics of imports and exports of nuclear fuels and basic materials which is set up by the Federal Authority for trade and industry for the Ministry of the Interior shows for 1981 (without taking the basic materials into account) a slight increase by 5% on the imports' side and also a slight increase by 10,5% on the exports' side. (orig./UA) [de

  16. Experience of development of the methods and equipment and the prospects for creation of WWER fuel examination stands

    International Nuclear Information System (INIS)

    Pavlov, S.; Smirnov, V.

    1998-01-01

    The report presents the basic methods and equipment developed for inspection of the fuel elements and fuel assemblies in the spent fuel pools. It considers their characteristics and results of the tests under laboratory and experimental fuel examination stand conditions. In particular, the following techniques are presented: visual inspection, measurement of the geometrical dimensions, definition of the form change in fuel assemblies and fuel elements, detection of the failed fuel elements, etc. The experience of the experimental fuel examination stand operation is generalized. The concept of the creation of the WWER-440 and WWER-1000 FA and FE inspection stands is presented. The concept is based on the modular principle which runs as follows. A set of the basic functional blocks is being developed based on which it is possible to make such a stand configuration which is necessary to fulfil the specific program of the examination at the particular nuclear power plant. (author)

  17. Innovative inert matrix-thoria fuels for in-reactor plutonium disposition

    International Nuclear Information System (INIS)

    Vettraino, F.; Padovani, E.; Luzzi, L.; Lombardi, C.; Thoresen, H.; Oberlander, B.; Iversen, G.; Espeland, M.

    1999-01-01

    The present leading option for plutonium disposition, either civilian or weapons Pu, is to burn it in LWRs after having converted it to MOX fuel. However, among the possible types of fuel which can be envisaged to burn plutonium in LWRs, innovative U-free fuels such as inert matrix and thoria fuel are novel concept in view of a more effective and ultimate solution from both security and safety standpoint. Inert matrix fuel is an non-fertile oxide fuel consisting of PuO 2 , either weapon-grade or reactor-grade, diluted in inert oxides such as for ex. stabilized ZrO 2 or MgAl 2 O 4 , its primary advantage consisting in no-production of new plutonium during irradiation, because it does not contain uranium (U-free fuel) whose U-238 isotope is the departure nuclide for breeding Pu-239. Some thoria addition in the matrix (thoria-doped fuel) may be required for coping with reactivity feedback needs. The full thoria-plutonia fuel though still a U-free variant cannot be defined non-fertile any more because the U-233 generation. The advantage of such a fuel option consisting basically on a remarkable already existing technological background and a potential acceleration in getting rid of the Pu stocks. All U-free fuels are envisaged to be operated under a once-through cycle scheme being the spent fuel outlooked to be sent directly to the final disposal in deep geological formations without requiring any further reprocessing treatment, thanks to the quality-poor residual Pu and a very high chemical stability under the current fuel reprocessing techniques. Besides, inert matrix-thoria fuel technology is suitable for in-reactor MAs transmutation. An additional interest in Th containing fuel refers to applicability in ADS, the innovative accelerated driven subcritical systems, specifically aimed at plutonium, minor actnides and long lived fission products transmutation in a Th-fuel cycle scheme which enables to avoid generations of new TRUs. A first common irradiation experiment

  18. Development of fuel performance and thermal hydraulic technology

    International Nuclear Information System (INIS)

    Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

    2000-03-01

    Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

  19. Development of the fabrication technology of the simulated DUPIC fuel

    International Nuclear Information System (INIS)

    Kang, Kweon Ho; Yang, M. S.; Bae, K. K. and others

    2000-06-01

    It is important to get basic data to analysis physical properties, behavior in reactor and performance of the DUPIC fuel because physical properties of the DUPIC fuel is different from the commercial UO 2 fuel. But what directly measures physical properties et al. of DUPIC fuel being resinterred simulated spent fuel through OREOX process is very difficult in laboratory owing to its high level radiation. Then fabrication of simulated DUPIC fuel is needed to measure its properties. In this study, processes on powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using simulated spent fuel are discribed. To fabricate simulated DUPIC fuel, the powder from 3 times OREOX and 5 times attrition milling simulated spent fuel is compacted with 1.3 ton/cm 2 . Pellets are sintered in 100% H 2 atmosphere over 10 h at 1800 deg C. Sintered densities of pellets are 10.2-10.5 g/cm 3

  20. Scientific issues in fuel behaviour

    International Nuclear Information System (INIS)

    1995-01-01

    The current limits on discharge burnup in today's nuclear power stations have proven the fuel to be very reliable in its performance, with a negligibly small rate of failure. However, for reasons of economy, there are moves to increase the fuel enrichment in order to extend both the cycle time and the discharge burnup. But, longer periods of irradiation cause increased microstructural changes in the fuel and cladding, implying a larger degradation of physical and mechanical properties. This degradation may well limit the plant life, hence the NSC concluded that it is of importance to develop a predictive capability of fuel behaviour at extended burnup. This can only be achieved through an improved understanding of the basic underlying phenomena of fuel behaviour. The Task Force on Scientific Issues Related to Fuel Behaviour of the NEA Nuclear Science Committee has identified the most important scientific issues on the subject and has assigned priorities. Modelling aspects are listed in Appendix A and discussed in Part II. In addition, quality assurance process for performing and evaluating new integral experiments is considered of special importance. Main activities on fuel behaviour modelling, as carried out in OECD Member countries and international organisations, are listed in Part III. The aim is to identify common interests, to establish current coverage of selected issues, and to avoid any duplication of efforts between international agencies. (author). refs., figs., tabs

  1. Fermentation, gasification and pyrolysis of carbonaceous residues towards usage in fuel cells

    International Nuclear Information System (INIS)

    Sequeira, C.A.C.; Brito, P.S.D.; Mota, A.F.; Carvalho, J.L.; Rodrigues, L.F.F.T.T.G.; Santos, D.M.F.; Barrio, D.B.; Justo, D.M.

    2007-01-01

    In this paper, the technologies of fermentation, gasification and pyrolysis of carbonaceous residues for the production of biohydrogen and other gaseous, liquid or solid fuels, are analysed. The energetic, economic and environmental advantages of linking these energy areas with the fuel cell engines are stressed. In addition, the current status of fuel cell technologies, namely their historic trends, basic electrode mechanisms, cell types, market drivers and leading issues, are reviewed

  2. Fact sheet on fuel manufacturing and utilization

    International Nuclear Information System (INIS)

    2006-01-01

    The Nuclear Fuel Cycle and Materials Section (NFCMS) supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle, provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection, supports the development fuel modeling expertise in Member States, covering both normal operation and postulated and severe accident conditions, provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation, supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, provides information and support research into the basic properties of fuel materials, including UO2, MOX, (Th, Pu)O2, (Th, U233)O2 fuels and zirconium alloy cladding and fuel assembly components and offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology and materials, economic and other aspects of nuclear fuel use (e.g. environmental impact). Recently NFCMS provided support to a Member State manufacturing Gadolinia doped fuel and provided in-mast sipping equipment to a Nuclear Power Plant to allow the determination of fuel failure. Member States interested in fuel performance and manufacture should contact the Technical Cooperation Department of the Agency and Member States interested in knowing more about the Agency's programme on source management should contact: C. Ganguly, Section Head, V. Inozemtsev, J. Killeen

  3. Fuel cells in transportation

    Energy Technology Data Exchange (ETDEWEB)

    Erdmann, G [Technische Univ., Berlin (Germany); Hoehlein, B [Research Center Juelich (Germany)

    1996-12-01

    A promising new power source for electric drive systems is the fuel cell technology with hydrogen as energy input. The worldwide fuel cell development concentrates on basic research efforts aiming at improving this new technology and at developing applications that might reach market maturity in the very near future. Due to the progress achieved, the interest is now steadily turning to the development of overall systems such as demonstration plants for different purposes: electricity generation, drive systems for road vehicles, ships and railroads. This paper does not present results concerning the market potential of fuel cells in transportation but rather addresses some questions and reflections that are subject to further research of both engineers and economists. Some joint effort of this research will be conducted under the umbrella of the IEA Implementing Agreement 026 - Annex X, but there is a lot more to be done in this challenging but also promising fields. (EG) 18 refs.

  4. Fast breeder fuel cycle

    International Nuclear Information System (INIS)

    1978-07-01

    This contribution is prepared for the answer to the questionnaire of working group 5, subgroup B. B.1. is the short review of the fast breeder fuel cycles based on the reference large commercial Japanese LMFBR. The LMFBRs are devided into two types. FBR-A is the reactor to be used before 2000, and its burnup and breeding ratio are relatively low. The reference fuel cycle requirement is calculated based on the FBR-A. FBR-B is the one to be used after 2000, and its burnup and breeding ratio are relatively high. B.2. is basic FBR fuel reprocessing scheme emphasizing the differences with LWR reprocessing. This scheme is based on the conceptual design and research and development work on the small scale LMFBR reprocessing facility of Japan. The facility adopts a conventional PUREX process except head end portions. The report also describes the effects of technical modifications of conventional reprocessing flow sheets, and the problems to be solved before the adoption of these alternatives

  5. Nuclear Fuel Cycle Analysis and Simulation Tool (FAST)

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Kwon, Eun Ha; Kim, Ho Dong

    2005-06-15

    This paper describes the Nuclear Fuel Cycle Analysis and Simulation Tool (FAST) which has been developed by the Korea Atomic Energy Research Institute (KAERI). Categorizing various mix of nuclear reactors and fuel cycles into 11 scenario groups, the FAST calculates all the required quantities for each nuclear fuel cycle component, such as mining, conversion, enrichment and fuel fabrication for each scenario. A major advantage of the FAST is that the code employs a MS Excel spread sheet with the Visual Basic Application, allowing users to manipulate it with ease. The speed of the calculation is also quick enough to make comparisons among different options in a considerably short time. This user-friendly simulation code is expected to be beneficial to further studies on the nuclear fuel cycle to find best options for the future all proliferation risk, environmental impact and economic costs considered.

  6. Information technology in fuel manufacturing

    International Nuclear Information System (INIS)

    Seshagiri Rao, G.R.; Arora, U.K.; Mohanty, Deepak; Siva Kumar, G.V.S.M.; Banerjee, P.K.

    2012-01-01

    NFC, Hyderabad is engaged in manufacturing of fuel assemblies required for Indian Nuclear Power Programme. During the manufacturing process, the basic Uranium Fuel and Zirconium alloy cladding tubes travels through several work centers, machines and exposes to various process parameters. For analyzing the fuel performance these parameters are indicators and is a requirement to record such history by both manufacturer and customer. NFC has planned to deploy Information Technology (IT) Systems from MDU/UOC Dissolution to Finished Fuel Assembly dispatch stage by using Radio Frequency IDentification (RFID)/Barcode Technologies. IT Systems are connected to electronic weigh balances to acquire material weight data automatically. The IT Systems are also designed to receive data from small Island Systems like Helium leak testing equipments. As a pilot project the system is initially implemented from empty tube Tray preparation stage to Fuel Assembly Packing and dispatch stage, containing about more than 14 processes. The system was built using open source technology platform and was deployed on a cost effective Hardware environment. The present paper describes the development process of the system, Implementation challenges faced and change management. The paper also discusses about fruits of implementation and productivity improvements. (author)

  7. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Rummery, T.E.; Rosinger, E.L.J.

    1984-12-01

    The Canadian Nuclear Fuel Waste Management Program is in the fourth year of a ten-year generic research and development phase. The objective of this phase of the program is to assess the basic safety and environmental aspects of the concept of isolating immobilized fuel waste by deep underground disposal in plutonic rock. The major scientific and engineering components of the program, namely immobilization studies, geoscience research, and environmental and safety assessment, are described. Program funding, scheduling and associated external review processes are briefly outlined

  8. Methods for estimating the reliability of the RBMK fuel assemblies and elements

    International Nuclear Information System (INIS)

    Klemin, A.I.; Sitkarev, A.G.

    1985-01-01

    Applied non-parametric methods for calculation of point and interval estimations for the basic nomenclature of reliability factors for the RBMK fuel assemblies and elements are described. As the fuel assembly and element reliability factors, the average lifetime is considered at a preset operating time up to unloading due to fuel burnout as well as the average lifetime at the reactor transient operation and at the steady-state fuel reloading mode of reactor operation. The formulae obtained are included into the special standardized engineering documentation

  9. Status of the Solid Oxide Fuel Cell Development at Topsoe Fuel Cell A/S and DTU Energy Conversion

    DEFF Research Database (Denmark)

    Christiansen, N.; Primdahl, S.; Wandel, Marie

    2013-01-01

    Many years of close collaboration between Topsoe Fuel Cell A/S (TOFC) and Risø (to day DTU Energy Conversion) on SOFC development have ensured an efficient transfer of SOFC basic know how to industrial technology. The SOFC development in the consortium includes material development...... and manufacturing of materials, cells and stacks based on state of the art as well as innovative strategies. Today TOFC provides the SOFC technology platform: Cells, stacks, integrated multi stack module and PowerCore units that integrate stack modules with hot fuel processing units for high electrical efficiency...

  10. Development and preliminary analyses of material balance evaluation model in nuclear fuel cycle

    International Nuclear Information System (INIS)

    Matsumura, Tetsuo

    1994-01-01

    Material balance evaluation model in nuclear fuel cycle has been developed using ORIGEN-2 code as basic engine. This model has feature of: It can treat more than 1000 nuclides including minor actinides and fission products. It has flexibility of modeling and graph output using a engineering work station. I made preliminary calculation of LWR fuel high burnup effect (reloading fuel average burnup of 60 GWd/t) on nuclear fuel cycle. The preliminary calculation shows LWR fuel high burnup has much effect on Japanese Pu balance problem. (author)

  11. The uranium-plutonium breeder reactor fuel cycle

    International Nuclear Information System (INIS)

    Salmon, A.; Allardice, R.H.

    1979-01-01

    All power-producing systems have an associated fuel cycle covering the history of the fuel from its source to its eventual sink. Most, if not all, of the processes of extraction, preparation, generation, reprocessing, waste treatment and transportation are involved. With thermal nuclear reactors more than one fuel cycle is possible, however it is probable that the uranium-plutonium fuel cycle will become predominant; in this cycle the fuel is mined, usually enriched, fabricated, used and then reprocessed. The useful components of the fuel, the uranium and the plutonium, are then available for further use, the waste products are treated and disposed of safely. This particular thermal reactor fuel cycle is essential if the fast breeder reactor (FBR) using plutonium as its major fuel is to be used in a power-producing system, because it provides the necessary initial plutonium to get the system started. In this paper the authors only consider the FBR using plutonium as its major fuel, at present it is the type envisaged in all, current national plans for FBR power systems. The corresponding fuel cycle, the uranium-plutonium breeder reactor fuel cycle, is basically the same as the thermal reactor fuel cycle - the fuel is used and then reprocessed to separate the useful components from the waste products, the useful uranium and plutonium are used again and the waste disposed of safely. However the details of the cycle are significantly different from those of the thermal reactor cycle. (Auth.)

  12. Using plasma-fuel systems at Eurasian coal-fired thermal power stations

    Science.gov (United States)

    Karpenko, E. I.; Karpenko, Yu. E.; Messerle, V. E.; Ustimenko, A. B.

    2009-06-01

    The development of plasma technology for igniting solid fuels at coal-fired thermal power stations in Russia, Kazakhstan, China, and other Eurasian countries is briefly reviewed. Basic layouts and technical and economic characteristics of plasma-fuel systems installed in different coal-fired boiles are considered together with some results from using these systems at coal-fired thermal power stations.

  13. Fuel Cell Handbook, Fifth Edition

    Energy Technology Data Exchange (ETDEWEB)

    Energy and Environmental Solutions

    2000-10-31

    Progress continues in fuel cell technology since the previous edition of the Fuel Cell Handbook was published in November 1998. Uppermost, polymer electrolyte fuel cells, molten carbonate fuel cells, and solid oxide fuel cells have been demonstrated at commercial size in power plants. The previously demonstrated phosphoric acid fuel cells have entered the marketplace with more than 220 power plants delivered. Highlighting this commercial entry, the phosphoric acid power plant fleet has demonstrated 95+% availability and several units have passed 40,000 hours of operation. One unit has operated over 49,000 hours. Early expectations of very low emissions and relatively high efficiencies have been met in power plants with each type of fuel cell. Fuel flexibility has been demonstrated using natural gas, propane, landfill gas, anaerobic digester gas, military logistic fuels, and coal gas, greatly expanding market opportunities. Transportation markets worldwide have shown remarkable interest in fuel cells; nearly every major vehicle manufacturer in the U.S., Europe, and the Far East is supporting development. This Handbook provides a foundation in fuel cells for persons wanting a better understanding of the technology, its benefits, and the systems issues that influence its application. Trends in technology are discussed, including next-generation concepts that promise ultrahigh efficiency and low cost, while providing exceptionally clean power plant systems. Section 1 summarizes fuel cell progress since the last edition and includes existing power plant nameplate data. Section 2 addresses the thermodynamics of fuel cells to provide an understanding of fuel cell operation at two levels (basic and advanced). Sections 3 through 8 describe the six major fuel cell types and their performance based on cell operating conditions. Alkaline and intermediate solid state fuel cells were added to this edition of the Handbook. New information indicates that manufacturers have stayed

  14. Basic plans on measures of mine site at the Ningyo-Toge Environmental Engineering Center

    International Nuclear Information System (INIS)

    2002-04-01

    At the Ningyo-Toge and its peripheries, there are some mine relating facilities and apparatuses finishing their actions such as wasted stones and slags accumulation sites, and so on formed by results of searching and mining works of uranium mine carried out from beginning of 1950s by the Nuclear Fuels Corporation and the Power Reactor and Nuclear Fuel Development Corporation, both of which are predecessors of the Japan Nuclear Cycle Development Institute (JNC). These facilities are, at present, adequately maintained and managed by the Ningyo-Toge Environmental Engineering Center, but as resource development of uranium was positioned to a disposal business on JNC, JNC has investigated on optimal measuring methods and testing plans to evaluate their safety under cooperation with other works of JNC, to summarize a draft of the basic plans on measures of mine site'. Here were described two drafts of the 'Basic plans on measures of mine site' summarized on concepts and indications of whole of measures of mine relating facilities sites and of the 'Proof test plan' summarized on testing plans containing concrete measures to obtain basic data and knowledge required for progressing the measures and a proof test. (G.K.)

  15. Choosing a spent fuel interim storage system

    International Nuclear Information System (INIS)

    Roland, V.; Hunter, I.

    2001-01-01

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

  16. 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

  17. As nuclear fuel bank project moves ahead, support for facility cannot falter

    Energy Technology Data Exchange (ETDEWEB)

    Shepherd, John [nuclear 24, Redditch (United Kingdom)

    2016-10-15

    During the summer 2016, the historic next steps were taken to establish an international nuclear fuel bank under the auspices of the International Atomic Energy Agency (IAEA). The 'bank', officially known as the IAEA Low Enriched Uranium (LEU) Storage Facility is scheduled to be ready for operations by this time next year. The key role of the fuel bank will be to hold a reserve of LEU, the basic ingredient of nuclear fuel.

  18. Gaseous fuel reactors for power systems

    International Nuclear Information System (INIS)

    Helmick, H.H.; Schwenk, F.C.

    1978-01-01

    The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems. Low power reactor experiments using uranium hexafluoride gas as fuel demonstrated performance in accordance with reactor physics predictions. The final phase of experimental activity now in progress is the fabrication and testing of a buffer gas vortex confinement system

  19. PULSTAR fuel, low enrichment, long lifetime, economical, proven

    International Nuclear Information System (INIS)

    Carter, Robert E.; Leonard, Bobby E.

    1993-01-01

    In 1962, the Western New York Research Center, Inc., located at the State University of New York at Buffalo, decided they had a need for a reactor with pulsing and high power steady state capabilities. Both General Atomic and the American Machine and Foundry Corporation (AMF) were contacted to ascertain if it were feasible to construct a dual purpose reactor of this type. The General Atomic proposal indicated the feasibility but would not warrant a steady state power of 2 MW with ultimate capability of 5 MW. AMF did provide a conceptual design for such a dual reactor, call the PULSTAR, and sufficient design information to confirm that the operating specifications could be met. The PULSTAR fuel consisted of 6 enrichment UO 2 sintered pellets in zircaloy tubes (pins) mounted in a x 5 array inside a fuel assembly. The fuel design was patterned after fuel that was under development for light water power reactors and that had been extensively tested under high power pulse conditions in the SPERT Test Reactor. The fuel assemblies are rectangular in a horizontal cross section, 315 inches by 2.74 inches, allowing for flat control blades to be inserted in the core grid arrangement. The active height of the core is approximately 24 inches. In the initial Buffalo AMF contract, a collaborative development agreement was signed in conjunction with agreement to construct the facility. After completion of the Buffalo PULSTAR Reactor, the PULSTAR fuel underwent an extensive test program which resulted in some minor changes in the basic design. In 1965, North Carolina State University contracted with AMF for the construction of a dual MW steady state (with ultimate capability of 5 MW and pulsing PULSTAR Research Reactor. Their fuel is identical to the Buffalo fuel except for having an enrichment of 4% U-235. This paper presented basic information about the characteristics and performance of the PULSTAR Research Reactor fuel. The following summarizes this information. The fuel is of

  20. PULSTAR fuel, low enrichment, long lifetime, economical, proven

    Energy Technology Data Exchange (ETDEWEB)

    Carter, Robert E; Leonard, Bobby E [Institute for Resource Management, Inc., Bethesda, MD (United States)

    1993-08-01

    In 1962, the Western New York Research Center, Inc., located at the State University of New York at Buffalo, decided they had a need for a reactor with pulsing and high power steady state capabilities. Both General Atomic and the American Machine and Foundry Corporation (AMF) were contacted to ascertain if it were feasible to construct a dual purpose reactor of this type. The General Atomic proposal indicated the feasibility but would not warrant a steady state power of 2 MW with ultimate capability of 5 MW. AMF did provide a conceptual design for such a dual reactor, call the PULSTAR, and sufficient design information to confirm that the operating specifications could be met. The PULSTAR fuel consisted of 6 enrichment UO{sub 2} sintered pellets in zircaloy tubes (pins) mounted in a x 5 array inside a fuel assembly. The fuel design was patterned after fuel that was under development for light water power reactors and that had been extensively tested under high power pulse conditions in the SPERT Test Reactor. The fuel assemblies are rectangular in a horizontal cross section, 315 inches by 2.74 inches, allowing for flat control blades to be inserted in the core grid arrangement. The active height of the core is approximately 24 inches. In the initial Buffalo AMF contract, a collaborative development agreement was signed in conjunction with agreement to construct the facility. After completion of the Buffalo PULSTAR Reactor, the PULSTAR fuel underwent an extensive test program which resulted in some minor changes in the basic design. In 1965, North Carolina State University contracted with AMF for the construction of a dual MW steady state (with ultimate capability of 5 MW and pulsing PULSTAR Research Reactor. Their fuel is identical to the Buffalo fuel except for having an enrichment of 4% U-235. This paper presented basic information about the characteristics and performance of the PULSTAR Research Reactor fuel. The following summarizes this information. The fuel

  1. Post irradiation examination of HANARO nucler mini-element fuel (metallographic and density test)

    International Nuclear Information System (INIS)

    Yoo, Byung Ok; Hong, K. P.; Park, D. G.; Choo, Y. S.; Baik, S. J.; Kim, K. H.; Kim, H. C.; Jung, Y. H.

    2001-05-01

    The post irradiation examination of a HANARO mini-element nuclear fuel, KH96C-004, was done in June 6, 2000. The purpose of this project is to evaluate the in-core performance and reliability of mini-element nuclear fuel for HANARO developed by the project T he Nuclear Fuel Material Development of Research Reactor . And, in order to examine the performance of mini-element nuclear fuel in normal output condition, the post irradiation examination of a nuclear fuel bundle composed by 6 mini nuclear fuel rods and 12 dummy fuel rods was performed. Based on these examination results, the safety and reliability of HANARO fuel and the basic data on the design of HANARO nuclear fuel can be ensured and obtained,

  2. Electricity generation from the mud by using microbial fuel cell

    Directory of Open Access Journals (Sweden)

    Idris Sitinoor Adeib

    2016-01-01

    Full Text Available Microbial fuel cells (MFCs is a bio-electrochemical device that harnesses the power of respiring microbes to convert organic substrates directly into electrical energy. This is achieved when bacteria transfer electrons to an electrode rather than directly to an electron acceptor. Their technical feasibility has recently been proven and there is great enthusiasm in the scientific community that MFCs could provide a source of “green electricity”. Microbial fuel cells work by allowing bacteria to do what they do best, oxidize and reduce organic molecules. Bacterial respiration is basically one big redox reaction in which electrons are being moved around. The objective is to generate electricity throughout the biochemical process using chemical waste basically sludge, via microbial fuel cells. The methodology includes collecting sludge from different locations, set up microbial fuel cells with the aid of salt bridge and observing the results in voltage measurement. The microbial fuel cells consist of two chambers, iron electrodes, copper wire, air pump (to increase the efficiency of electron transfer, water, sludge and salt bridge. After several observations, it is seen that this MFC can achieve up until 202 milivolts (0.202volts with the presence of air pump. It is proven through the experiments that sludge from different locations gives different results in term of the voltage measurement. This is basically because in different locations of sludge contain different type and amount of nutrients to provide the growth of bacteria. Apart from that, salt bridge also play an important role in order to transport the proton from cathode to anode. A longer salt bridge will give a higher voltage compared to a short salt bridge. On the other hand, the limitations that this experiment facing is the voltage that being produced did not last long as the bacteria activity slows down gradually and the voltage produced are not really great in amount. Lastly to

  3. Quality analysis in pressurized water reactor fuel

    International Nuclear Information System (INIS)

    Darolles, J.F.

    1975-01-01

    An integrated system which has been set up to administrate and analyze the quality is described. This system is in actual operation. The basic principles for quality analysis system are traceability, i.e., identification, location and history of fuel components and quality evaluation during manufacturing. The quality analysis system operates in the following areas: data recording and transmission, data processing, quality file generation. The interest of such a system may be noted particularly in manufacturing, for the constitution of quality files, the design of products and the processing of data from irradiated fuel assemblies [fr

  4. The JRC-ITU approach to the safety of advanced nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Fanghaenel, T.; Rondinella, V.V.; Somers, J.; Konings, R.; Erdmann, N.; Uffelen, P. van; Glatz, J.P. [European Commission, Joint Research Centre - JRC, Institute for Transuranium Elements - ITU, Postfach 2340, 76125 Karlsruhe (Germany)

    2013-07-01

    The JRC-ITU safety studies of advanced fuels and cycles adopt two main axes. First the full exploitation of still available and highly relevant knowledge and samples from past fuel preparation and irradiation campaigns (complementing the limited number of ongoing programmes). Secondly, the shift of focus from simple property measurement towards the understanding of basic mechanisms determining property evolution and behaviour of fuel compounds during normal, off-normal and accident conditions. The final objective of the second axis is the determination of predictive tools applicable to systems and conditions different from those from which they were derived. State of the art experimental facilities, extensive networks of partnerships and collaboration with other organizations worldwide, and a developing programme for training and education are essential in this approach. This strategy has been implemented through various programs and projects. The SUPERFACT programme constitutes the main body of existing knowledge on the behavior in-pile of MOX fuel containing minor actinides. It encompassed all steps of a closed fuel cycle. Another international project investigating the safety of a closed cycle is METAPHIX. In this case a U-Pu19-Zr10 metal alloy containing Np, Am and Cm constitutes the fuel. 9 test pins have been prepared and irradiated. In addition to the PIE (Post Irradiation Examination), pyrometallurgical separation of the irradiated fuel has been performed, to demonstrate all the steps of a multiple recycling closed cycle and characterize their safety relevant aspects. Basic studies like thermodynamic fuel properties, fuel-cladding-coolant interactions have also been carried out at JRC-ITU.

  5. Determination of Basic Structure-Property Relations for Processing and Modeling in Advanced Nuclear Fuel: Microstructure Evolution and Mechanical Properties

    International Nuclear Information System (INIS)

    Wheeler, Kirk; Parra, Manuel; Peralta, Pedro

    2009-01-01

    The project objective is to study structure-property relations in solid solutions of nitrides and oxides with surrogate elements to simulate the behavior of fuels of inert matrix fuels of interest to the Advanced Fuel Cycle Initiative (AFCI), with emphasis in zirconium-based materials. Work with actual fuels will be carried out in parallel in collaboration with Los Alamos National Laboratory (LANL). Three key aspects will be explored: microstructure characterization through measurement of global texture evolution and local crystallographic variations using Electron Backscattering Diffraction (EBSD); determination of mechanical properties, including fracture toughness, quasi-static compression strength, and hardness, as functions of load and temperature, and, finally, development of structure-property relations to describe mechanical behavior of the fuels based on experimental data. Materials tested will be characterized to identify the mechanisms of deformation and fracture and their relationship to microstructure and its evolution. New aspects of this research are the inclusion of crystallographic information into the evaluation of fuel performance and the incorporation of statistical variations of microstructural variables into simplified models of mechanical behavior of fuels that account explicitly for these variations. The work is expected to provide insight into processing conditions leading to better fuel performance and structural reliability during manufacturing and service, as well as providing a simplified testing model for future fuel production

  6. Metallic fission product releases from HTR-spherical fuel elements

    International Nuclear Information System (INIS)

    Helmbold, M.; Amian, W.; Stoever, D.; Hecker, R.

    1978-01-01

    Fission product releases from fuel determines to a large extent the feasibility of a special reactor concept. Basic data describing the diffusion behaviour from coated particle fuel are presented concerning isotopes Cs 137 , Sr 90 and Agsup(110m). Taking into account these data for typical 3000MWth plants release calculations are performed. Sensitive release parameters could be defined and the results show low release figures for all the considered reactor concepts. (author)

  7. C-5M Fuel Efficiency Through MFOQA Data Analysis

    Science.gov (United States)

    2015-03-26

    then contribute to more accurate fuel loading and more efficient fleet fuel usage. Flight Operations Quality Assurance (FOQA) data uses the quick access ...into the ranges in Table 2. Visual Basic ( VBA ) code was written to quickly parse an entire mission (one of the thirty samples) into usable cruise...segments within the Altitude ranges of Table 2. The logic for focusing upon stable cruise flight segments was outlined in Chapter II. The VBA code

  8. The impact of catalytic materials on fuel reformulation

    Energy Technology Data Exchange (ETDEWEB)

    Rossini, Stefano [Snamprogetti, S. Donato Milanese, Milan (Italy)

    2003-01-15

    Fuel reformulation has been seeded by the growing consciousness of the potential damages mankind was causing to the ecosystem and to itself. Fuel reformulation means that fuels are defined on a chemical composition base with additional engine-technology related standards rather than on pure performance bases. These standards, which are getting more and more stringent, can be met by different leverages, mainly catalysts and processes operating conditions.This survey reviews the contribution of catalytic materials to the production of cleaner fuel components through some significant examples selected from scientific and technical literature. Having described the trends in automotive fuels quality, production of gasoline and diesel pool components is discussed relating the required properties to the material active site configuration, i.e. acidity/basicity, structural parameters, physical constraints. While distinctions are made between pathways leading to gasoline and those leading to diesel, sulfur removal is faced on a more generalized approach.

  9. Decarbonising the energy intensive basic materials industry through electrification – Implications for future EU electricity demand

    International Nuclear Information System (INIS)

    Lechtenböhmer, Stefan; Nilsson, Lars J.; Åhman, Max; Schneider, Clemens

    2016-01-01

    The need for deep decarbonisation in the energy intensive basic materials industry is increasingly recognised. In light of the vast future potential for renewable electricity the implications of electrifying the production of basic materials in the European Union is explored in a what-if thought-experiment. Production of steel, cement, glass, lime, petrochemicals, chlorine and ammonia required 125 TW-hours of electricity and 851 TW-hours of fossil fuels for energetic purposes and 671 TW-hours of fossil fuels as feedstock in 2010. The resulting carbon dioxide emissions were equivalent to 9% of total greenhouse gas emissions in EU28. A complete shift of the energy demand as well as the resource base of feedstocks to electricity would result in an electricity demand of 1713 TW-hours about 1200 TW-hours of which would be for producing hydrogen and hydrocarbons for feedstock and energy purposes. With increased material efficiency and some share of bio-based materials and biofuels the electricity demand can be much lower. Our analysis suggest that electrification of basic materials production is technically possible but could have major implications on how the industry and the electric systems interact. It also entails substantial changes in relative prices for electricity and hydrocarbon fuels. - Highlights: • Energy intensive basic materials industry has a high share in EU greenhouse gas emissions. • Decarbonising these industries is very important, but still relatively unexplored. • Electrification is possible regarding renewable energy resources and technologies. • Combination with energy and materials efficiency, biofuels and CCS is crucial. • Electrification needs very high amounts of electricity and strong policies.

  10. Basic program of atomic energy development and utilization for fiscal 1981

    International Nuclear Information System (INIS)

    1981-01-01

    Nuclear power generation is capable of supplying large quantity of energy as the core of petroleum substitutes. Besides its costs are low, it can contribute in number of ways, such as the suppression of price rise and the stabilization of international balance of payments. Its development and utilization are the important aspects of the energy policy of Japan. In the promotion of atomic energy development, securing its safety is the foremost prerequisite. Meanwhile, the nuclear fuel cycle must be established as early as possible, concerning such as the securing of uranium resources, the domestic production of enriched uranium and the establishment of domestic fuel reprocessing. The basic program in fiscal 1981 is described as follows: the strengthening of the safety measures, the promotion of nuclear power generation, the establishment of the nuclear fuel cycle, the research on nuclear fusion, and so on. (J.P.N.)

  11. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Kabele, T.J.

    1979-09-01

    To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems.

  12. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

    International Nuclear Information System (INIS)

    Schneider, K.J.; Kabele, T.J.

    1979-09-01

    To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems

  13. Fuel cycles using adulterated plutonium

    International Nuclear Information System (INIS)

    Brooksbank, R.E.; Bigelow, J.E.; Campbell, D.O.; Kitts, F.G.; Lindauer, R.B.

    1978-01-01

    Adjustments in the U-Pu fuel cycle necessitated by decisions made to improve the nonproliferation objectives of the US are examined. The uranium-based fuel cycle, using bred plutonium to provide the fissile enrichment, is the fuel system with the highest degree of commercial development at the present time. However, because purified plutonium can be used in weapons, this fuel cycle is potentially vulnerable to diversion of that plutonium. It does appear that there are technologically sound ways in which the plutonium might be adulterated by admixture with 238 U and/or radioisotopes, and maintained in that state throughout the fuel cycle, so that the likelihood of a successful diversion is small. Adulteration of the plutonium in this manner would have relatively little effect on the operations of existing or planned reactors. Studies now in progress should show within a year or two whether the less expensive coprocessing scheme would provide adequate protection (coupled perhaps with elaborate conventional safeguards procedures) or if the more expensive spiked fuel cycle is needed as in the proposed civex pocess. If the latter is the case, it will be further necessary to determine the optimum spiking level, which could vary as much as a factor of a billion. A very basic question hangs on these determinations: What is to be the nature of the recycle fuel fabrication facilities. If the hot, fully remote fuel fabrication is required, then a great deal of further development work will be required to make the full cycle fully commercial

  14. Casting of metallic fuel containing minor actinide additions

    International Nuclear Information System (INIS)

    Trybus, C.L.; Henslee, S.P.; Sanecki, J.E.

    1992-01-01

    A significant attribute of the Integral Fast Reactor (IFR) concept is the transmutation of long-lived minor actinide fission products. These isotopes require isolation for thousands of years, and if they could be removed from the waste, disposal problems would be reduced. The IFR utilizes pyroprocessing of metallic fuel to separate auranium, plutonium, and the minor actinides from nonfissionable constituents. These materials are reintroduced into the fuel and reirradiated. Spent IFR fuel is expected to contain low levels of americium, neptunium, and curium because the hard neutron spectrum should transmute these isotopes as they are produced. This opens the possibility of using an IFR to trnasmute minor actinide waste from conventional light water reactors (LWRs). A standard IFR fuel is based on the alloy U-20% Pu-10% Zr (in weight percent). A metallic fuel system eases the requirements for reprocessing methods and enables the minor actinide metals to be incorporated into the fuel with simple modifications to the basic fuel casting process. In this paper, the authors report the initial casting experience with minor actinide element addition to an IFR U-Pu-Zr metallic fuel

  15. The DUPIC alternative for backend fuel cycle

    International Nuclear Information System (INIS)

    Lee, J.S.; Yang, M.S.; Park, H.S.; Boczar, P.; Sullivan, J.; Gadsby, R.D.

    1997-01-01

    The DUPIC fuel cycle was conceived as an alternative to the conventional fuel cycle backed options, with a view to multiple benefits expectable from burning spent PWR fuel again in CANDU reactors. It is based on the basic idea that the bulk of spent PWR fuel can be directly refabricated into a reusable fuel for CANDU of which high efficiency in neutron utilization would exhaustively burn the fissile remnants in the spent PWR fuel to a level below that of natural uranium. Such ''burn again'' strategy of the DUPIC fuel cycle implies that the spent PWR fuel will become CANDU fuel of higher burnup with relevant benefits such as spent PWR fuel disposition, saving of natural uranium fuel, etc. A salient feature of the DUPIC fuel cycle is neither the fissile content nor the bulk radioactivity is separated from the DUPIC mass flow which must be contained and shielded all along the cycle. This feature can be considered as a factor of proliferation resistance by deterrence against access to sensitive materials. It means also the requirement for remote systems technologies for DUPIC fuel operation. The conflicting aspects between better safeguardability and harder engineering problems of the radioactive fuel operation may be the important reason why the decades' old concept, since INFCE, of ''hot'' fuel cycle has not been pursued with much progress. In this context, the DUPIC fuel cycle could be a live example for development of proliferation resistant fuel cycle. As the DUPIC fuel cycle looks for synergism of fuel linkage from PWR to CANDU (or in broader sense LWR to HWR), Korea occupies a best position for DUPIC exercise with her unique strategy of reactor mix of both reactor types. But the DUPIC benefits can be extended to global bonus, expectable from successful development of the technology. (author)

  16. Pyrometallurgical separation processes of radionuclides contained in the irradiated nuclear fuel

    International Nuclear Information System (INIS)

    De Cordoba, Guadalupe; Caravaca, Concha; Quinones, Javier; Gonzalez de la Huebra, Angel

    2005-01-01

    Faced with the new options for the high level waste management, the ''Partitioning and Transmutation (P and T)'' of the radio nuclides contained in the irradiated nuclear fuel appear as a promising option from different points of view, such as environmental risk, radiotoxic inventory reduction, economic, etc.. The present work is part of a research project called ''PYROREP'' of the 5th FWP of the EU that studied the feasibility of the actinide separation from the rest of fission products contained in the irradiated nuclear fuel by pyrometallurgical processes with the aim of their transmutation. In order to design these processes it is necessary to determine basic thermodynamic and kinetic data of the radionuclides contained in the nuclear fuel in molten salt media. The electrochemical study of uranium, samarium and molybdenum in the eutectic melt LiCl - KCl has been performed at a tungsten electrode in the temperature range of 450 - 600 deg C in order to obtain these basic properties. (Author)

  17. Proposal for basic safety requirements regarding the disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    1980-04-01

    A working group commissioned to prepare proposals for basic safety requirements for the storage and transport of radioactive waste prepared its report to the Danish Agency of Environmental Protection. The proposals include: radiation protection requirements, requirements concerning the properties of high-level waste units, the geological conditions of the waste disposal location, the supervision of waste disposal areas. The proposed primary requirements for safety evaluation of the disposal of high-level waste in deep geological formations are of a general nature, not being tied to specific assumptions regarding the waste itself, the geological and other conditions at the place of disposal, and the technical methods of disposal. It was impossible to test the proposals for requirements on a working repository. As no country has, to the knowledge of the working group, actually disposed of hifg-level radioactive waste or approved of plans for such disposal. Methods for evaluating the suitability of geological formations for waste disposal, and background material concerning the preparation of these proposals for basic safety requirements relating to radiation, waste handling and geological conditions are reviewed. Appended to the report is a description of the phases of the fuel cycle that are related to the storage of spent fuel and the disposal of high-level reprocessing waste in a salt formation. It should be noted that the proposals of the working group are not limited to the disposal of reprocessed fuel, but also include the direct disposal of spent fuel as well as disposal in geological formations other than salt. (EG)

  18. Overview of MOX fuel fabrication achievements

    International Nuclear Information System (INIS)

    Bairiot, H.; Vliet, J. van; Chiarelli, G.; Edwards, J.; Nagai, S.H.; Reshetnikov, F.

    2000-01-01

    Such overview having been adequately covered in an OECD/NEA publication providing the situation as of end 1994, this paper is mainly devoted to an update as of end 1998. The Belgian plant, Belgonucleaire/Dessel, is now dedicated exclusively to the fabrication of MOX fuel and has operated consistently around its nameplate capacity (35tHM/a) through the 1990s involving a large variety of PWR and BWR fuels. The two French plants have also achieved routine operation during the 1990s. CFCa, historically the largest FBR MOX fuel manufacturer, is utilizing the genuine COCA process for that type of fuel and the MIMAS process for LWR fuel: a nominal capacity (40 tHM/a) has been gradually approached. MELOX has operated at 100 tHM/a, as defined in the operating licence granted originally. The British plant, MDF/Sellafield with 8tHM/a nameplate capacity is devoted to fuel and has manufactured several small fabrication campaigns. In Japan, JNC operates three facilities located at Tokai: PFDF, devoted to basic research and fabrication of test fuels, PFFF/ATR line, for the fabrication of Fugen fuel and of corresponding fuel for the critical facility DCA, and PFPF for the fabrication of FBR fuel. In Russia, fabrication techniques have been developed to fuel four BN-800 FBRs contemplated to be constructed and be fuelled with the civilian Pu stockpile. Two demonstration facilities Paket (Mayak) and RIAR (Dimitrovgrad) fabricated respectively pellet and vipac type FBR MOX fuel for BR-5, BOR-60, BN-350 and BN-600. The paper includes a brief description of each of the fabrication routes mentioned, as well as the production of respectively LWR and FBR MOX fuel in each fabrication facility, since the start-up of the plant, since 1 January 1993 and since 1 January 1998 up to 31 December 1998. (author)

  19. Status of the Canadian Nuclear Fuel Waste Management Program

    International Nuclear Information System (INIS)

    Lyon, R.B.

    1985-10-01

    The Canadian Nuclear Fuel Waste Management Program is in the fifth year of a ten-year generic research and development phase. The major objective of this phase of the program is to assess the basic safety and environmental aspects of the concept of isolating immobilized fuel waste by deep underground disposal in plutonic rock. The major scientific and engineering components of the program, namely immobilization studies, geoscience research, and environmental and safety assessment, are well established

  20. Two-dimensional analytical model of a proton exchange membrane fuel cell

    International Nuclear Information System (INIS)

    Liu, Jia Xing; Guo, Hang; Ye, Fang; Ma, Chong Fang

    2017-01-01

    In this study, a two-dimensional full cell analytical model of a proton exchange membrane fuel cell is developed. The analytical model describes electrochemical reactions on the anode and cathode catalyst layer, reactants diffusion in the gas diffusion layer, and gases flow in the gas channel, etc. The analytical solution is derived according to the basic physical equations. The performance predicted by the model is in good agreement with the experimental data. The results show that the polarization mainly occurs in the cathode side of the proton exchange membrane fuel cell. The anodic overpotential cannot be neglected. The hydrogen and oxygen concentrations decrease along the channel flow direction. The hydrogen and oxygen concentrations in the catalyst layer decrease with the current density. As predicted by the model, concentration polarization mainly occurs in the cathode side. - Highlights: • A 2D full cell analytical model of a proton exchange membrane fuel cell is developed. • The analytical solution is deduced according to the basic equations. • The anode overpotential is not so small that it cannot be neglected. • Species concentration distributions in the fuel cell is obtained and analyzed.

  1. Nagra technical report 14-02, geological basics - Dossier III - Long-term geological developments

    International Nuclear Information System (INIS)

    Schnellmann, M.; Madritsch, H.

    2014-01-01

    This dossier is the third of a series of eight reports concerning the safety and technical aspects of locations for the disposal of radioactive wastes in Switzerland. Dossier III takes a look at long-term geological developments. Developments in the topography and river networks of northern Switzerland over the past five million years are looked at. Data and information derived from high-resolution models and compilations of gravel deposition, glacier developments and moraines are reviewed. Tectonic developments, seismological aspects and erosion are discussed. Their consequences for the long-term geological developments in the proposed depository areas are looked at

  2. Tensile mechanical properties of U3Si2-Al fuel plate

    International Nuclear Information System (INIS)

    Xu Yong; Hu Huawei; Zhuang Hongquan; Wang Xishu

    2003-01-01

    The fuel plate made of fuel meat, with the U 3 Si 2 -Al dispersion fuel center, and 6061 Al alloy cladding, is a new kind of fuel used in research reactors. The mechanical property data of the fuel meat is the basic data in the design of fuel group, but the mechanical property of this fuel meat has not been studied all over the world till now. In this paper, the mechanical properties of U 3 Si 2 -Al fuel meats of different sizes used in research reactors are investigated and analyzed, and at the same time the carrying capacity of tensile in different directions are also compared. In order to get more knowledge about the mechanical properties of the fuel meat, the tensile experiment has been carried out repeatedly. Considering the lower ratio of elongation and the brittleness, the microscope has been used to examine the zone of fracture after tensile test. (authors)

  3. Fuel cycle and waste newsletter, Vol. 4, No. 1, April 2008

    International Nuclear Information System (INIS)

    2008-04-01

    This issue of the Fuel Cycle and Waste Newsletter presents the International Decommissioning Network, the cooperation between INPRO (the International Project on Innovative Nuclear Reactors and Fuel Cycles) and NEFW (IAEA's Division of Nuclear Fuel Cycle and Waste Technology), the policies and strategies for spent fuel and radioactive waste management, recent developments of decommissioning waste, integrated approach to decommissioning and environmental remediation, CEG Workshop, repatriation of sealed sources in Latin America, the technical working Group on research reactors (TWGRR), an update on research reactor networks, Atominstitut Vienna, modernization and refurbishment of research reactors, a new CRP on innovative methods in research reactor analysis, management of damaged spent nuclear fuel, influence of high-burnup UOX and MOX water reactor fuel on spent fuel management, a new CRP on improvement in the computer code modelling of high burnup nuclear fuel (FUMEX-3), reuse options for reprocessed uranium (RepU), a basic fact-book on coated particle fuel, recent publications and upcoming meetings

  4. Fuel assembly assessment from CVD image analysis: A feasibility study

    International Nuclear Information System (INIS)

    Lindsay, C.S.; Lindblad, T.

    1997-05-01

    The Swedish Nuclear Inspectorate commissioned a feasibility study of automatic assessment of fuel assemblies from images obtained with the digital Cerenkov viewing device currently in development. The goal is to assist the IAEA inspectors in evaluating the fuel since they typically have only a few seconds to inspect an assembly. We report results here in two main areas: Investigation of basic image processing and recognition techniques needed to enhance the images and find the assembly in the image; Study of the properties of the distributions of light from the assemblies to determine whether they provide unique signatures for different burn-up and cooling times for real fuel or indicate presence of non-fuel. 8 refs, 27 figs

  5. Reactor-specific spent fuel discharge projections: 1985 to 2020

    International Nuclear Information System (INIS)

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

    1986-09-01

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

  6. Current concept of the nuclear fuel cycle in the Czech Republic

    International Nuclear Information System (INIS)

    Priman, V.; Vesely, P.; Sedina, M.

    2001-01-01

    The article provides an overview of the major activities within the nuclear fuel cycle as practised by the Czech utility CEZ, a. s. Efforts in the fuel cycle practice are focused on the implementation of standard business principles with CEZ's contractual suppliers; application of new legislation and international design standards; and on achieving a higher technical standard of nuclear fuel and its better overall utilisation by nuclear power plants. Business activities in the procurement of nuclear materials and their diversification are discussed. The last part of the article, which is devoted to the fuel cycle back-end, describes the adopted strategy of extended spent fuel storage associated with a postponement of the final disposal. The basic principles of the CEZ's fuel cycle back-end strategy are also summarised. (author)

  7. Fuel radial design using Path Relinking

    International Nuclear Information System (INIS)

    Campos S, Y.

    2007-01-01

    The present work shows the obtained results when implementing the combinatory optimization technique well-known as Path Re linking (Re-linkage of Trajectories), to the problem of the radial design of nuclear fuel assemblies, for boiling water reactors (BWR Boiling Water Reactor by its initials in English), this type of reactors is those that are used in the Laguna Verde Nucleo electric Central, Veracruz. As in any other electric power generation plant of that make use of some fuel to produce heat and that it needs each certain time (from 12 to 14 months) to make a supply of the same one, because this it wears away or it burns, in the nucleolectric plants to this activity is denominated fuel reload. In this reload different activities intervene, among those which its highlight the radial and axial designs of fuel assemblies, the patterns of control rods and the multi cycles study, each one of these stages with their own complexity. This work was limited to study in independent form the radial design, without considering the other activities. These phases are basic for the fuel reload design and of reactor operation strategies. (Author)

  8. Thermogravimetric analysis of fuel film evaporation

    Institute of Scientific and Technical Information of China (English)

    HU Zongjie; LI Liguang; YU Shui

    2006-01-01

    Thermogravimetric analysis (TGA) was compared with the petrochemical distillation measurement method to better understand the characteristics of fuel film evaporation at different wall tem- peratures. The film evaporation characteristics of 90# gasoline, 93# gasoline and 0# diesel with different initial thicknesses were investigated at different environmental fluxes and heating rates. The influences of heating rate, film thickness and environmental flux on fuel film evaporation for these fuels were found. The results showed that the environmental conditions in TGA were similar to those for fuel films in the internal combustion engines, so data from TGA were suitable for the analysis of fuel film evaporation. TGA could simulate the key influencing factors for fuel film evaporation and could investigate the basic quantificational effect of heating rate and film thickness. To get a rapid and sufficient fuel film evaporation, sufficiently high wall temperature is necessary. Evaporation time decreases at a high heating rate and thin film thickness, and intense gas flow is important to promoting fuel film evaporation. Data from TGA at a heating rate of 100℃/min are fit to analyze the diesel film evaporation during cold-start and warming-up. Due to the tense molecular interactions, the evaporation sequence could not be strictly divided according to the boiling points of each component for multicomponent dissolved mixture during the quick evaporation process, and the heavier components could vaporize before reaching their boiling points. The 0# diesel film would fully evaporate when the wall temperature is beyond 250℃.

  9. Cost aspects of the research reactor fuel cycle

    International Nuclear Information System (INIS)

    2010-01-01

    Research reactors have made valuable contributions to the development of nuclear power, basic science, materials development, radioisotope production for medicine and industry, and education and training. In doing so, they have provided an invaluable service to humanity. Research reactors are expected to make important contributions in the coming decades to further development of the peaceful uses of nuclear technology, in particular for advanced nuclear fission reactors and fuel cycles, fusion, high energy physics, basic research, materials science, nuclear medicine, and biological sciences. However, in the context of decreased public sector support, research reactors are increasingly faced with financial constraints. It is therefore of great importance that their operations are based on a sound understanding of the costs of the complete research reactor fuel cycle, and that they are managed according to sound financial and economic principles. This publication is targeted at individuals and organizations involved with research reactor operations, with the aim of providing both information and an analytical framework for assessing and determining the cost structure of fuel cycle related activities. Efficient management of fuel cycle expenditures is an important component in developing strategies for sustainable future operation of a research reactor. The elements of the fuel cycle are presented with a description of how they can affect the cost efficient operation of a research reactor. A systematic review of fuel cycle choices is particularly important when a new reactor is being planned or when an existing reactor is facing major changes in its fuel cycle structure, for example because of conversion of the core from high enriched uranium (HEU) to low enriched uranium (LEU) fuel, or the changes in spent fuel management provision. Review and optimization of fuel cycle issues is also recommended for existing research reactors, even in cases where research reactor

  10. Reactor-specific spent fuel discharge projections, 1987-2020

    International Nuclear Information System (INIS)

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

    1988-03-01

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

  11. Dry Refabrication Technology Development of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Lee, Jung Won; Park, G. I.; Park, C. J.

    2010-04-01

    Key technical data on advanced nuclear fuel cycle technology development for the spent fuel recycling have been produced in this study. In the frame work of DUPIC, dry process oxide products fabrication, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remote modulated welding equipment has been designed and fabricated. In the area of advanced pre-treatment process development, a rotary-type oxidizer and spherical particle fabrication process were developed by using SIMFUEL and off-gas treatment technology and zircalloy tube treatment technology were studied. In the area of the property characteristics of dry process products, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data

  12. Nagra technical report 14-02, geological basics - Dossier VII - Usage conflicts

    International Nuclear Information System (INIS)

    Gautschi, A.; Becker, J.; Traber, D.; Leu, W.

    2014-01-01

    This dossier is the seventh of a series of eight reports concerning the safety and technical aspects of locations for the disposal of radioactive wastes in Switzerland. It discusses possible conflicts with respect to the use of rock strata below or above the proposed host rock layers. Possible usage could include the extraction of salt, coal or other hydrocarbons. Other possible conflicting uses include the mining of stone, ores and minerals as well as the extraction of mineral water and thermal water. The construction of deep boreholes, for example for geothermal probes, could also cause conflicts with any nuclear waste depositories. The storage of natural gas or carbon sequestration, however, is not considered likely

  13. 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

  14. Use of non-proliferation fuel cycles in the HTGR

    International Nuclear Information System (INIS)

    Baxter, A.M.; Merrill, M.H.; Dahlberg, R.C.

    1978-10-01

    All high-temperature gas-cooled reactors (HTGRs) built or designed to date utilize a uranium-thorium fuel cycle (HEU/Th) in which fully-enriched uranium (93% U-235) is the initial fuel and thorium is the fertile material. The U-233 produced from the thorium is recycled in subsequent loadings to reduce U-235 makeup requirements. However, the recent interest in proliferation-proof fuel cycles for fission reactors has prompted a review and evaluation of possible alternate cycles in the HTGR. This report discusses these alternate fuel cycles, defines those considered usable in an HTGR core, summarizes their advantages and disadvantages, and briefly describes the effect on core design of the most important cycles. Examples from design studies are also given. These studies show that the flexibility afforded by the HTGR coated-particle fuel design allows a variety of alternative cycles, each having special advantages and attractions under different circumstances. Moreover, these alternate cycles can all use the same fuel block, core layout, control scheme, and basic fuel zoning concept

  15. Internal combustion engines for alcohol motor fuels: a compilation of background technical information

    Energy Technology Data Exchange (ETDEWEB)

    Blaser, Richard

    1980-11-01

    This compilation, a draft training manual containing technical background information on internal combustion engines and alcohol motor fuel technologies, is presented in 3 parts. The first is a compilation of facts from the state of the art on internal combustion engine fuels and their characteristics and requisites and provides an overview of fuel sources, fuels technology and future projections for availability and alternatives. Part two compiles facts about alcohol chemistry, alcohol identification, production, and use, examines ethanol as spirit and as fuel, and provides an overview of modern evaluation of alcohols as motor fuels and of the characteristics of alcohol fuels. The final section compiles cross references on the handling and combustion of fuels for I.C. engines, presents basic evaluations of events leading to the use of alcohols as motor fuels, reviews current applications of alcohols as motor fuels, describes the formulation of alcohol fuels for engines and engine and fuel handling hardware modifications for using alcohol fuels, and introduces the multifuel engines concept. (LCL)

  16. Development of technology of handling with ampoule source of ionizing radiation

    International Nuclear Information System (INIS)

    Urenskij, N.A.; Boltovskij, S.A.; Ganzha, V.V.; Shuklin, G.S.

    2000-01-01

    It is reported that in Kazakhstan at different enterprises there are more than 80,000 ampoule source of ionizing radiation (ASIR) and about 20,000 of them should be disposed. In 1995 the first phase of depository for spent ASIR was putted into operation. The depository is placed on 'Baikal-1' stand complex. The first phase of depository consists of following buildings and equipment: 2 special equipped motorcar; temporary depository; special washing machine; radiation protected chamber, intermediate depository; lift crane with set of special grips; protective container; jackets; cassette-holders; constant depository. Design of the depository, its equipment, system of dosimetric control and qualified personnel provide the security of operation and exclude the environment contamination

  17. Hydrogen fuel cell power system

    International Nuclear Information System (INIS)

    Lam, A.W.

    2004-01-01

    'Full text:' Batteries are typically a necessary and prime component of any DC power system, providing a source of on-demand stored energy with proven reliability. The integration of batteries and basic fuel cells for mobile and stationary utility applications poses a new challenge. For high value applications, the specification and operating requirements for this hybrid module differ from conventional requirements as the module must withstand extreme weather conditions and provide extreme reliability. As an electric utility company, BCHydro has embarked in the development and application of a Hydrogen Fuel Cell Power Supply (HFCPS) for field trial. A Proton Exchange Membrane (PEM)- type fuel cell including power electronic modules are mounted in a standard 19-inch rack that provides 48V, 24V, 12V DC and 120V AC outputs. The hydrogen supply consists of hydrogen bottles and regulating devices to provide a continuous fuel source to the power modules. Many tests and evaluations have been done to ensure the HFCPS package is robust and suitable for electric utility grade operation. A field trial demonstrating this standalone system addressed reliability, durability, and installation concerns as well as developed the overall system operating procedures. (author)

  18. Methanol supply issues for alternative fuels demonstration programs

    International Nuclear Information System (INIS)

    Teague, J.M.; Koyama, K.K.

    1995-01-01

    This paper surveys issues affecting the supply of fuel-grade methanol for the California Energy Commission's alternative fuels demonstration programs and operations by other public agencies such as transit and school districts. Establishing stable and reasonably priced sources of methanol (in particular) and of alternative fuels generally is essential to their demonstration and commercialization. Development both of vehicle technologies and of fuel supply and distribution are complementary and must proceed in parallel. However, the sequence of scaling up supply and distribution is not necessarily smooth; achievement of volume thresholds in demand and through-put of alternative fuels are marked by different kinds of challenges. Four basic conditions should be met in establishing a fuel supply: (1) it must be price competitive with petroleum-based fuels, at least when accounting for environmental and performance benefits; (2) bulk supply must meet volumes required at each phase; necessitating resilience among suppliers and a means of designating priority for high value users; (3) distribution systems must be reliable, comporting with end users' operational schedules; (4) volatility in prices to the end user for the fuel must be minimal. Current and projected fuel volumes appear to be insufficient to induce necessary economies of scale in production and distribution for fuel use. Despite their benefits, existing programs will suffer absent measures to secure economical fuel supplies. One solution is to develop sources that are dedicated to fuel markets and located within the end-use region

  19. Development of the fabrication technology of the simulated fuel-I, 15,000MWd/tU

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kweon Ho; Kim, D. J.; Kim, H. S.; Lee, J. W.; Yang, M. S

    2001-04-01

    It is important to get basic data to analysis physical properties, behavior in reactor and performance of the DUPIC fuel because physical properties, fission gas release, grain growth and et al. of the DUPIC fuel is different from the commercial UO2 fuel. But what directly measures physical properties et al. of DUPIC fuel being resinterred simulated spent fuel through OREOX process is very difficult in laboratory owing to its high level radiation. Then fabrication of simulated DUPIC fuel is needed to measure its properties. In this study, the sintering characterization of wet milled powder for 24 hours to fabricate 15,000MWd/tU equivalent burnup simulated fuel.

  20. Car buyers and fuel economy?

    International Nuclear Information System (INIS)

    Turrentine, Thomas S.; Kurani, Kenneth S.

    2007-01-01

    This research is designed to help researchers and policy makers ground their work in the reality of how US consumers are thinking and behaving with respect to automotive fuel economy. Our data are from semi-structured interviews with 57 households across nine lifestyle 'sectors.' We found no household that analyzed their fuel costs in a systematic way in their automobile or gasoline purchases. Almost none of these households track gasoline costs over time or consider them explicitly in household budgets. These households may know the cost of their last tank of gasoline and the unit price of gasoline on that day, but this accurate information is rapidly forgotten and replaced by typical information. One effect of this lack of knowledge and information is that when consumers buy a vehicle, they do not have the basic building blocks of knowledge assumed by the model of economically rational decision-making, and they make large errors estimating gasoline costs and savings over time. Moreover, we find that consumer value for fuel economy is not only about private cost savings. Fuel economy can be a symbolic value as well, for example among drivers who view resource conservation or thrift as important values to communicate. Consumers also assign non-monetary meaning to fuel prices, for example seeing rising prices as evidence of conspiracy. This research suggests that consumer responses to fuel economy technology and changes in fuel prices are more complex than economic assumptions suggest. The US Department of Energy and the Energy Foundation supported this research. The authors are solely responsible for the content and conclusions presented

  1. Effect of power variations across a fuel bundle and within a fuel element on fuel centerline temperature in PHWR bundles in uncrept and crept pressure tubes

    International Nuclear Information System (INIS)

    Onder, E.N.; Roubtsov, D.; Rao, Y.F.; Wilhelm, B.

    2017-01-01

    Highlights: • Pressure tube creep effect on fuel pin power and temperatures was investigated. • Noticeable effects were observed for 5.1% crept pressure tube. • Bundle eccentricity effect on power variations was insignificant for uncrept channels. • Difference of 112 °C was observed between top & bottom elements in 5.1% crept channel. • Not discernible fission gas release was expected with temperature difference of 112 °C. - Abstract: The neutron flux and fission power profiles through a fuel bundle and across a fuel element are important aspects of nuclear fuel analysis in multi-scale/multi-physics modelling of Pressurized Heavy Water Reactors (PHWRs) with advanced fuel bundles. Fuel channels in many existing PHWRs are horizontal. With ageing, pressure tubes creep and fuel bundles in these pressure tubes are eccentrically located, which results in an asymmetric coolant flow distribution between the top and bottom of the fuel bundles. The diametral change of the pressure tube due to creep is not constant along the fuel channel; it reaches a maximum in the vicinity of the maximum neutron flux location. The cross-sectional asymmetric positioning of fuel bundles in a crept pressure tube contributes to an asymmetric power distribution within a ring of fuel elements. Modern reactor physics lattice codes (such as WIMS-AECL) are capable of predicting the details of power distribution from basic principles. Thermalhydraulics subchannel codes (such as ASSERT-PV) use models to describe inhomogeneous power distribution within and across fuel elements (e.g., flux tilt model, different powers in different ring elements, or radial power profiles). In this work, physics and thermalhydraulics codes are applied to quantify the effect of eccentricity of a fuel bundle on power variations across it and within a fuel element, and ultimately on the fuel temperature distribution and fuel centerline temperature, which is one of the indicators of fuel performance under normal

  2. Operation of spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

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

  3. Development of fast reactor metal fuels containing minor actinides

    International Nuclear Information System (INIS)

    Ohta, Hirokazu; Ogata, Takanari; Kurata, Masaki; Koyama, Tadafumi; Papaioannou, Dimitrios; Glatz, Jean-Paul; Rondinella, Vincenzo V.

    2011-01-01

    Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phenix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU. (author)

  4. Actual Status of CAREM-25 Fuel Element Development

    International Nuclear Information System (INIS)

    Perez, Edmundo

    2000-01-01

    In the frame of the CAREM Project, under Cnea s Reactor and Nuclear Plants Program, the Nuclear Fuel Thematic Area is one among others on which the project is organized. In this area, the primary objective to reach is to actualize the mechanical fuel element and reactivity control designs, taking in account the recents conceptual and engineering modifications introduced in the reactor, and ending with a consolidated conceptual and basic development.In order to reach these objectives, it is presented the way on which the area was organized, the participating working groups, the task required, the personnel involucrated, the grade of global development reached in the areas of engineering, developments, fabrication and essays of design verification, and the found difficulties, the tasks under ejecution, just finished and necessaries to fulfill completely the objectives. Finally, it is possible to say that due to the work realized, the conceptual design of both components is finished and the basic design is under development

  5. An investigation on technical feasibilities of fuel cycle for high temperature gas-cooled reactor (Case study)

    International Nuclear Information System (INIS)

    Sumita, Junya; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sawa, Kazuhiro

    2008-03-01

    In accordance with the basic policy of effectively using nuclear fuel resources, the FBR cycle, one of the most possible fuel cycle in the future, will be adapted after plu-thermal program by LWR in Japanese nuclear cycle plan. In this paper, a case study of technical investigation of HTGR fuel cycle based on HTGR fuel cycle proposed to adapt to Japanese nuclear fuel cycle plan were carried out from the viewpoint of effective utilization of uranium, fabrication technologies of MOX fuel, reprocessing technologies, amount of interim storage of HTGR fuel and graphite waste. As a result, the fuel cycle for HTGR is expected to be possible technically. (author)

  6. Field experience of new nuclear fuel types on the Kola NPP

    International Nuclear Information System (INIS)

    Adeev, V.; Burlov, S.; Panov, A.; Saprykin, V.

    2008-01-01

    without regenerated uranium was started. In this paper schemes of profiling 2-nd generation fuel assemblies and CR fuel followers, even and odd core loading patterns of a steady-state fuel cycle, the basic characteristics of the realized core loading patterns and the basic results of operation are presented. The specified characteristics of a steady-state fuel cycle with assemblies of 2-nd generation and a comparison between calculated and experimental shutdown efficiency of the unit 4 Kola NPP are also given

  7. A study on KMRR utilization for fuel development

    International Nuclear Information System (INIS)

    Kang, Young Hwan; Ryu, Woo Seog; Park, Ji Yeon; Joo, Kee Nam; Park, Jong Man; Park, Se Jin

    1991-01-01

    The most effective utilization scheme of the KMRR was studied in the field of nuclear fuel development through reviewing literatural documents on irradiation facilities and in-pile test. It is suggested that the KMRR should be used for verification tests of advanced fuels and for power ramping / cycling tests of fuel rods. In addition, the characterization tests for fuel development and the basic material research should be also performed. In-pile loops for fuel verification and/or power ramping / cycling tests are proposed to be installed in advance, and capsules are necessary for power ramping / cycling tests, fuel characterization tests and / or material tests. Instrumentation technologies for thermocouple, SPND (Self-Powered Neutron Detector) and pressure transducer, and the in-situ dimensional measuring systems have to be developed to obtain the useful and various results from irradiation tests in the KMRR. A mock-up test rod for characterizing fuel thermal response was manufactured and the related technologies as well as the design specification were set up. An equipment for microdrilling and grooving of fuel pellets and an apparatus for diffusion-bonding between zircaloy-4 and stainless steel were made. A study to verify the integrity of test rod weldments is presented using out-of pile corrosion test. (Author)

  8. Status and prospects of safety research about fuel cycle facilities in France

    International Nuclear Information System (INIS)

    Auchere, H.; Mercier, J.P.

    1996-01-01

    Although there is a good knowledge of the risks and no major accident occurred in France, as in other OECD countries, it remains useful to complete basic knowledge and to allow the quality of fuel cycle plants safety assessments to be improved further, particularly in countries equipped with a 'complete' nuclear fuel cycle (France, Japan and U.K.). The scope of the current and future IPSN ('Institut de Protection et de Surete Nucleaire': institute for protection and nuclear safety) research deals with the whole fuel cycle. The overview presented here in NUCEF'95 symposium contains a number of specific themes, some of which have already been started. Successful conclusion of the safety researches will allow the IPSN to have a more precise understanding about specific phenomena and notably to replace 'engineer judgements', though they may be based on a lot of experience and competence, by more scientifically established basic data. (J.P.N.)

  9. State of the art: Multi-fuel reformers for automotive fuel cell applications. Problem identification and research needs

    Energy Technology Data Exchange (ETDEWEB)

    Westerholm, R. [Stockholm Univ. (Sweden). Dept. of Analytical Chemistry; Pettersson, L.J. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Chemical Engineering and Technology

    1999-12-01

    On an assignment from the Transport and Communications Research Board (KFB) a literature study and a study trip to the USA and Great Britain have been performed. The literature study and the study trip was made during late spring and autumn 1999.The purpose of the project was to collect available information about the chemical composition of the product gas from a multi-fuel reformer for a fuel cell vehicle. It was furthermore to identify problems and research needs. The report recommends directions for future major research efforts. The results of the literature study and the study trip led to the following general conclusions: With the technology available today it does not seem feasible to develop a highly efficient and reliable multi-fuel reformer for automotive applications, i. e. for applications where all types of fuels ranging from natural gas to heavy diesel fuels can be used. The potential for developing a durable and reliable system is considerably higher if dedicated fuel reformers are used.The authors propose that petroleum-derived fuels should be designed for potential use in mobile fuel cell applications. In the present literature survey and the site visit discussions we found that there are relatively low emissions from fuel cell engines compared to internal combustion engines. However, the major research work on reformers/fuel cells have been performed during steady-state operation. Emissions during start-up, shutdown and transient operation are basically unknown and must be investigated in more detail. The conclusions and findings in this report are based on open/available information, such as discussions at site visits, reports, scientific publications and symposium proceedings.

  10. Study on the nitride fuel fabrication for FBR cycle (1)

    International Nuclear Information System (INIS)

    Shinkai, Yasuo; Ono, Kiyoshi; Tanaka, Kenya

    2002-07-01

    In the phase-II of JNC's 'Feasibility Study on Commercialized Fuel Reactor Cycle System (the F/S)', the nitride fuels are selected as candidate for fuels for heavy metal cooled reactor, gas cooled reactor, and small scale reactor. In particular, the coated fuel particles are a promising concept for gas cooled reactor. In addition, it is necessary to study in detail the application possibility of pellet nitride fuel and vibration compaction nitride fuel for heavy metal cooled reactor and small scale reactor in the phase-II. In 2001, we studied more about additional equipments for the nitride fuel fabrication in processes from gelation to carbothermic reduction in the vibration compaction method. The result of reevaluation of off-gas mass flow around carbothermic reduction equipment in the palletizing method, showed that quantity of off-gas flow reduced and its reduction led the operation cost to decrease. We studied the possibility of fabrication of large size particles in the coated fuel particles for helium gas cooled reactor and we made basic technical issues clear. (author)

  11. An Evaluation of a Fission Product Inventory for CANDU Fuels

    International Nuclear Information System (INIS)

    Jung, Jong Yeob; Park, Joo Hwan

    2007-01-01

    Fission products are released by two processes when a single channel accident occurs. One is a 'prompt release' and the other is a 'delayed release'. Prompt release assumes that the gap inventory of the fuel elements is released by a fuel element failure at the time of an accident. Delayed release assumes that the inventories within the grain or at the grain boundary are released after a accident due to a diffusion through grains, an oxidation of the fuel and an interaction between the fuel and the Zircaloy sheath. Therefore, the calculation of a fission product inventory and its distribution in a fuel during a normal operating is the starting point for the assessment of a fission product release for single channel accidents. In this report, the fission product inventories and their distributions within s fuel under a normal operating condition are evaluated for three types of CANDU fuels such as the 37 element fuel, CANFLEX-NU and CANFLEX-RU fuel bundles in the 'limiting channel'. To accomplish the above mentioned purposes, the basic power histories for each type of CANDU fuel were produced and the fission product inventories were calculated by using the ELESTRES code

  12. Mechanical energy release and fuel fragmentation in high energy deposition into fuel under a reactivity initiated accident condition

    International Nuclear Information System (INIS)

    Tsuruta, Takaharu; Saito, Shinzo; Ochiai, Masaaki

    1985-01-01

    The fuel fragmentation is one of important subjects to be studied, since it is one of basic processes of molten fuel-coolant interaction (MFCI) and it has not yet been made clear enough. Accordingly, UO 2 fuel fragmentation was studied in the NSRR experiments simulating a reactivity initiated accident (RIA). As results of the experiments, the distribution of the size of fuel fragments was obtained and the mechanism of fuel fragmentation was discussed as described below. It was revealed that the distribution was well displayed in the form of logarithmic Rosin-Rammler's distribution law. It was shown that the conversion ratio from thermal energy to mechanical in the experiment was in inverse propotion to the volume-surface mean diameter defined as a ratio of the total volume of fragments to the total surface. Consequently, it was confirmed that the mean diameter was proper as an index for the degree of the fuel fragmentation. It was also pointed out that the Weber-type hydraulic instability model for fragmentation was consistent with the experimental results. The mechanism of the fuel fragmentation is understood as follows. Cladding tube is ruptured due to the increase in rod pressure when fuel is molten, and then molten fuel spouts through the openings in the form of jet. As a result of molten fuel spouting, fuel is fragmented by the Weber-type of hydraulic instability. The model well explains the effects of experimental parameters as heat deposition, subcooling of cooling water and capsule diameter, on the fuel fragmentation. According to the model, fuel fragments have to be spherical. There were many spherical particles which had hollow and burst crack. This may be due to internal burst during solidification process. The items which should be studied further are also described in the end of this report. (author)

  13. New UO2 fuel studies

    International Nuclear Information System (INIS)

    Dehaudt, P.; Lemaignan, C.; Caillot, L.; Mocellin, A.; Eminet, G.

    1998-01-01

    With improved UO 2 fuels, compared with the current PWR, one would enable to: retain the fission products, rise higher burn-ups and deliver the designed power in reactor for longer times, limit the pellet cladding interaction effects by easier deformation at high temperatures. Specific studies are made in each field to understand the basic mechanisms responsible for these improvements. Four programs on new UO 2 fuels are underway in the laboratory: advanced microstructure fuels (doped fuels), fuels containing Er 2 O 3 a burnable absorber, fuels with improved caesium retention, composite fuels. The advanced microstructure UO 2 fuels have special features such as: high grain sizes to lengthen the fission gas diffusion paths, intragranular precipitates as fission gas atoms pinning sites, intergranular silica based viscoplastic phases to improve the creep properties. The grain size growth can be obtained with a long time annealing or with corundum type oxide additives partly soluble in the UO 2 lattice. The amount of doping element compared with its solubility limit and the sintering conditions allows to obtain oxide or metallic precipitates. The fuels containing Er 2 O 3 as a burnable absorber are under irradiation in the TANOX device at the present time. Specific sintering conditions are required to improve the erbium solubility in UO 2 and to reach standard or large grain sizes. The improved caesium retention fuels are doped with SiO 2 +A1 2 O 3 or SiO 2 +ZrO 2 additives which may form stable compounds with the Cs element in accidental conditions. The composite fuels are made of UO 2 particles of about 100 μm in size dispersed in a molybdenum metallic (CERMET) or MgA1 2 O 4 ceramic (CERCER) matrix. The CERMET has a considerably higher thermal conductivity and remains ''cold'' during irradiation. The concept of double barrier (matrix+fuel) against fission products is verified for the CERMET fuel. A thermal analysis of all the irradiated rods shows that the thermal

  14. Basic plans of atomic energy development and utilization for fiscal 1978

    International Nuclear Information System (INIS)

    1978-01-01

    The Government has promoted the development and utilization of atomic energy as one of the most important measures for energy supplies. In Japan, due to the unrest concerning safety of nuclear power, siting of nuclear power plants is difficult, thereby the nuclear power generation program is delayed. Then, in major research and development projects such as those of uranium enrichment, fast breeder reactors, an advanced thermal reactor and nuclear fusion, while the remarkable results are being accumulated, the practical aspects are in need of positive governmental measures. Under this situation, the long range program of atomic energy development and utilization is being revised. For the fiscal year 1978 (from April, 1978 to March, 1979), based on the revision, the basic plans are presented, first, the basic policy, and second, the practical measures: strengthening of the safety measures; establishment of the nuclear fuel cycle; development of the new types of power reactors; promotion of the basic researches; securing of the people's understanding and cooperation. (Mori, K

  15. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1983-07-01

    This publication contains basically three kinds of information: routes approved by the Commission for the shipment of irradiated reactor fuel, information regarding any safeguards-significant incidents which have been reported to occur during shipments along such routes, and cumulative amounts of material shipped

  16. Plutonium bearing oxide fuels for recycling in thermal reactors and fast breeder reactors

    International Nuclear Information System (INIS)

    Cunningham, G.W.

    1977-01-01

    Programs carried out in the past two decades have established the technical feasibility of using plutonium as a fuel material in both water-cooled power reactors and sodium-cooled fast breeder reactors. The problem facing the technical community is basically one of demonstrating plutonium fuel recycle under strict conditions of public safety, accountability, personnel exposure, waste management, transportation and diversion or theft which are still evolving. In this paper only technical and economic aspects of high volume production and the demonstration program required are discussed. This paper discusses the role of mixed oxide fuels in light water reactors and the objectives of the LMFBR required for continual growth of nuclear power during the next century. The results of studies showing the impact of using plutonium on uranium requirements, power costs, and the market share of nuclear power are presented. The influence of doubling time and the introduction date of LMFBRs on the benefits to be derived by its commercial use are discussed. Advanced fuel development programs scoped to meet future commerical LMFBR fuel requirements are described. Programs designed to provide the basic technology required for using plutonium fuels in a manner which will satisfy all requirements for public acceptance are described. Included are the high exposure plutonium fabrication development program centered around the High Performance Fuels Laboratory being built at the Hanford Engineering Development Laboratory and the program to confirm the technology required for the production of mixed oxide fuels for light water reactors which is being coordinated by Savannah River Laboratories

  17. Biomass energy - large potential in North-West Russia

    International Nuclear Information System (INIS)

    Borchsenius, Hans

    2000-01-01

    Changing from oil or coal to bio fuel is a high priority in all European countries. The potential for such a transition is largest in North-West Russia, where several factors point to biomass energy: large bio fuel resources, large need for heating because of the cold climate, and almost 100% coverage of district heating. Here, the largest continuous coniferous forest in Europe supplies the raw material for a considerable forest industry, including some of the biggest sawmills and paper- and cellulose factories in the world. The fraction of the timber that cannot go into this production is suitable as bio fuel. About 15% of the raw material in this industry is bark and sawdust which can be used for energy production. In addition, 10% of the biomass of the trees remains on the forest floor as twigs, treetops etc. If all this sawdust and felling waste was used to replace heating oil, the corresponding reduction of CO2 emission would amount to 25 mill m3 per year. The forest industry in Russia is currently in full production, and an increasing mass of sawdust and wood waste is accumulating in depositories that cover larger and larger areas. Depositories are often set on fire to keep down the masses; at the same time, the district heating plants are fired with expensive oil or coal. This paradoxical situation is due to the economical crises in the 1990s. Neither private companies nor the local governments could invest in bio fueled boilers. Bio fuel projects are cost-effective and easy to document and perfectly suitable for joint implementations under the Kyoto Protocol

  18. A LMFBR for thorium utilization and for the U233/Th fuel rods specification

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Dias, A.F.

    1982-01-01

    The use of U 233 /Th as fuel in the middle part of LMFBR core and the Pu/U in the external part of the core, are proposed. The basic neutronic and safety characteristics and the specifications of fuel rods to be used in the internal core, are presented. (E.G.) [pt

  19. Cooperative Russian-French experiment on plutonium-enriched fuels for fast burner reactor

    International Nuclear Information System (INIS)

    Zabud'ko, L.M.; Kurina, I.A.; Men'shikova, T.S.; Rogozkin, B.D.; Maershin, A.A.; Langi, A.; Pillon, S.

    2001-01-01

    Various kinds of nuclear fuels with an increased plutonium content are under study according to the program including three stages: fabrication, irradiation in BOR-60 reactor, post-irradiation examination. Flowsheets for fabricating pelletized and vibrocompacted fuels of UPu 0.45 O 2 , UPu 0.45 N, UPu 0.6 N, PuN + ZrN, PuO 2 + MgO are presented along with basic fuel properties. The irradiation of oxide fuel is carried out in an individual irradiation device at rated maximum temperature of the fuel at the beginning of irradiation equal to 2100 deg C. The irradiation of nitride fuel and the fuel based on inert matrices is performed in the other device with the aim of limitation of maximum temperature by the value of 1550 deg C. The duration of irradiation for all fuel types constitutes 750 EFPD. Fuel element charge in Bor-60 reactor core was realized in 2000 [ru

  20. A review on the status of development in thorium-based nuclear fuels

    International Nuclear Information System (INIS)

    Lee, Young Woo; Na, S. H.; Lee, Y. W.; Kim, H. S.; Kim, S. H.; Joung, C.Y.

    2000-02-01

    Thorium as an alternative nuclear energy source had been widely investigated in the 1950s-1960s because it is more abundant than uranium, but the studies of thorium nuclear fuel cycle were discontinued by political and economic reasons in the 1970s. Recently, however, renewed interest was vested in thorium-based nuclear fuel cycle because it may generate less long-lived minor actinides and has a lower radiotoxicity of high level wastes after reprocessing compared with the thorium fuel cycle. In this state-of the art report, thorium-based nuclear cycle. In this state-of the art report, thorium-based nuclear fuel cycle and fuel fabrication processes developed so far with different reactor types are reviewed and analyzed to establish basic technologies of thorium fuel fabrication which could meet our situation. (author)

  1. Models for MOX fuel behaviour. A selective review

    International Nuclear Information System (INIS)

    Massih, Ali R.

    2006-01-01

    This report reviews the basic physical properties of light water reactor mixed-oxide (MOX) fuel comprising nuclear characteristics, thermal properties such as melting temperature, thermal conductivity, thermal expansion, and heat capacity, and compares these with properties of conventional UO 2 fuel. These properties are generally well understood for MOX fuel and are well described by appropriate models developed for engineering analysis. Moreover, certain modelling approaches of MOX fuel in-reactor behaviour, regarding densification, swelling, fission product gas release, helium release, fuel creep and grain growth, are evaluated and compared with the models for UO 2 . In MOX fuel the presence of plutonium rich agglomerates adds to the complexity of fuel behaviour on the micro scale. In addition, we survey the recent fuel performance experience and post irradiation examinations on several types of MOX fuel types. We discuss the data from these examinations, regarding densification, swelling, fission product gas release and the evolution of the microstructure during irradiation. The results of our review indicate that in general MOX fuel has a higher fission gas release and helium release than UO 2 fuel. Part of this increase is due to the higher operating temperatures of MOX fuel relative to UO 2 fuel due to the lower thermal conductivity of MOX material. But this effect by itself seems to be insufficient to make for the difference in the observed fission gas release of UO 2 vs. MOX fuel. Furthermore, the irradiation induced creep rate of MOX fuel is higher than that of UO 2 . This effect can reduce the pellet-clad interaction intensity in fuel rods. Finally, we suggest that certain physical based approaches discussed in the report are implemented in the fuel performance code to account for the behaviour of MOX fuel during irradiation

  2. Models for MOX fuel behaviour. A selective review

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2006-12-15

    This report reviews the basic physical properties of light water reactor mixed-oxide (MOX) fuel comprising nuclear characteristics, thermal properties such as melting temperature, thermal conductivity, thermal expansion, and heat capacity, and compares these with properties of conventional UO{sub 2} fuel. These properties are generally well understood for MOX fuel and are well described by appropriate models developed for engineering analysis. Moreover, certain modelling approaches of MOX fuel in-reactor behaviour, regarding densification, swelling, fission product gas release, helium release, fuel creep and grain growth, are evaluated and compared with the models for UO{sub 2}. In MOX fuel the presence of plutonium rich agglomerates adds to the complexity of fuel behaviour on the micro scale. In addition, we survey the recent fuel performance experience and post irradiation examinations on several types of MOX fuel types. We discuss the data from these examinations, regarding densification, swelling, fission product gas release and the evolution of the microstructure during irradiation. The results of our review indicate that in general MOX fuel has a higher fission gas release and helium release than UO{sub 2} fuel. Part of this increase is due to the higher operating temperatures of MOX fuel relative to UO{sub 2} fuel due to the lower thermal conductivity of MOX material. But this effect by itself seems to be insufficient to make for the difference in the observed fission gas release of UO{sub 2} vs. MOX fuel. Furthermore, the irradiation induced creep rate of MOX fuel is higher than that of UO{sub 2}. This effect can reduce the pellet-clad interaction intensity in fuel rods. Finally, we suggest that certain physical based approaches discussed in the report are implemented in the fuel performance code to account for the behaviour of MOX fuel during irradiation.

  3. Proceedings of the third specialist meeting on sodium/fuel interaction in fast reactors

    International Nuclear Information System (INIS)

    1976-01-01

    This specialist meeting, sponsored by the OECD-NEA and organized by the Power Reactor and Nuclear Fuel Development Corporation, was attended by 56 delegates from 6 countries and the CEC (Commission of the European Communities). The purpose of the meeting was to bring together and discuss in depth the Fuel-Sodium Interaction, a phenomenon of major importance in the assessment of the Hypothetical Core Disruptive Accident in the Liquid Metal Fast Breeder Reactor. The meeting was essentially a follow-up of an earlier meeting held at Ispra in December 1973. In all, 29 papers were presented, covering the following topics: 1. Current perspective on sodium-fuel interaction in LMFBR safety; 2. Basic experimental and theoretical studies including other materials; 3. In-pile and out-of-pile experimental studies on sodium-fuel interaction; 4. Theoretical models for the interpretation of experiments and for application to reactor situations. The meeting is considered useful in narrowing down the chain of events necessary to get energetic interaction, large work potential, but many points are being clarified on the gap between the basic vapor explosions and the real fuel sodium interactions in the HCDA scenario of LMFBR. Finally another meeting of the same nature as this one has been recommended

  4. Radioactive waste management of experimental DUPIC fuel fabrication process

    International Nuclear Information System (INIS)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Hong, K. P.

    2001-01-01

    The concept of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) is a dry processing technology to manufacture CANDU compatible DUPIC fuel from spent PWR fuel material. Real spent PWR fuel was used in IMEF M6 hot cell to carry out DUPIC experiment. Afterwards, about 200 kg-U of spent PWR fuel is supposed to be used till 2006. This study has been conducted in some hot cells of PIEF and M6 cell of IMEF. There are various forms of nuclear material such as rod cut, powder, green pellet, sintered pellet, fabrication debris, fuel rod, fuel bundle, sample, and process waste produced from various manufacturing experiment of DUPIC fuel. After completing test, the above nuclear wastes and test equipment etc. will be classified as radioactive waste, transferred to storage facility and managed rigorously according to domestic and international laws until the final management policy is determined. It is desirable to review management options in advance for radioactive waste generated from manufacturing experiment of DUPIC nuclear fuel as well as residual nuclear material and dismantled equipment. This paper includes basic plan for DUPIC radwaste, arising source and estimated amount of radioactive waste, waste classification and packing, transport cask, transport procedures

  5. High temperature fuel cell with ceria-based solid electrolyte

    International Nuclear Information System (INIS)

    Arai, H.; Eguchi, K.; Yahiro, H.; Baba, Y.

    1987-01-01

    Cation-doped ceria is investigated as an electrolyte for the solid oxide fuel cell. As for application to the fuel cells, the electrolyte are desired to have high ionic conductivity in deriving a large electrical power. A series of cation-doped ceria has higher ionic conductivity than zirconia-based oxides. In the present study, the basic electrochemical properties of cation-doped ceria were studied in relation to the application of fuel cells. The performance of fuel cell with yttria-doped ceria electrolyte was evaluated. Ceria-based oxides were prepared by calcination of oxide mixtures of the components or calcination of co-precipitated hydroxide mixtures from the metal nitrate solution. The oxide mixtures thus obtained were sintered at 1650 0 C for 15 hr in air into disks. Ionic transference number, t/sub i/, was estimated from emf of oxygen concentration cell. Electrical conductivities were measured by dc-4 probe method by varying the oxygen partial pressure. The fuel cell was operated by oxygen and hydrogen

  6. Standard guide for drying behavior of spent nuclear fuel

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This guide is organized to discuss the three major components of significance in the drying behavior of spent nuclear fuel: evaluating the need for drying, drying spent nuclear fuel, and confirmation of adequate dryness. 1.1.1 The guide addresses drying methods and their limitations in drying spent nuclear fuels that have been in storage at water pools. The guide discusses sources and forms of water that remain in SNF, its container, or both, after the drying process and discusses the importance and potential effects they may have on fuel integrity, and container materials. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments. 1.1.2 The basic issue in drying is to determine how dry the SNF must be in order to prevent issues with fuel retrievability, container pressurization, or container corrosion. Adequate d...

  7. Thermal-hydraulic Parameters of WWER-440 Fuel Performance

    International Nuclear Information System (INIS)

    Kacmar, M.

    1999-01-01

    In this lecture the thermodynamic characteristics of reactor core V230, V213 (design concept and the distribution of coolant flow inside the reactor vessel and basic T-H parameters), allowed reactor regimes and reactor core flow factor as well as thermodynamic parameters and impact to fuel leakage are presented

  8. Removal of Basic Nitrogen Compounds from Fuel Oil with [Hnmp]H2PO4 Ionic Liquid

    Directory of Open Access Journals (Sweden)

    Z. Zhou

    2017-04-01

    Full Text Available Ionic liquid (IL N-methyl pyrrolidone dihydrogen phosphate ([Hnmp]H2PO4 was synthesized and its structure was characterized with FT-IR spectroscopy and 1H NMR. The denitrogenation performance of the ionic liquid was investigated using Fushun shale diesel oil that included 0.52 w% basic nitrogen as feedstock. Experiment results showed that under the operating conditions with temperature of 30 °C, 1:7 (w/w IL: oil, reaction time of 20 min, and settling time of 2 h, the ionic liquid exhibited good denitrogenation performance achieving 86.27 % basic N-extraction efficiency and the yield of refined diesel oil can reach more than 90 %. In addition, the basic N-removal efficiency can still reach 54 % during four recycles of the ionic liquid.

  9. Synergistic Smart Fuel For Microstructure Mediated Measurements

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; Dale K. Kotter; Steven L. Garrett; Randall A. Ali

    2013-07-01

    Advancing the Nuclear Fuel Cycle and Next Generation Nuclear Power Plants requires enhancing our basic understanding of fuel and materials behavior under irradiation. The two most significant issues limiting the effectiveness and lifespan of the fuel are the loss of thermal conductivity of the fuel and the mechanical strength of both fuel and cladding. The core of a nuclear reactor presents an extremely harsh and challenging environment for both sensors and telemetry due to elevated temperatures and large fluxes of energetic and ionizing particles from radioactive decay processes. The majority of measurements are made in reactors using “radiation hardened” sensors and materials. A different approach has been pursued in this research that exploits high temperatures and materials that are robust with respect to ionizing radiation. This synergistically designed thermoacoustic sensor will be self-powered, wireless, and provide telemetry. The novel sensor will be able to provide reactor process information even if external electrical power and communication are unavailable. In addition, the form-factor for the sensor is identical to the existing fuel rods within reactors and contains no moving parts. Results from initial proof of concept experiments designed to characterize porosity, surface properties and monitor gas composition will be discussed.

  10. Report of the collaboration project for research and development of sphere-pac fuel among JNC-PSI-NRG (1). Planning, fuel design, pin fabrication

    International Nuclear Information System (INIS)

    Morihira, Masayuki; Ozawa, Takayuki; Tomita, Yutaka; Suzuki, Masahiro; Kihara, Yoshiyuki; Shigetome, Yoshiaki; Kohno, Shusaku

    2004-07-01

    The collaboration project concerning sphere-pac fuel among JNC, Swiss PSI (Paul Scherrer Institut) and Dutch NRG (Nuclear Research and Consultancy Group) is in progress. Final target of the project is comparative irradiation tests of sphere-pac fuel in the HFR (High Flux Reactor) in Petten in the Netherlands with pellet type fuel and vipack fuel. Total 16 fuel segments (8 pins) of these three types of fuel are planned to be irradiated. Two sphere-pac fuel segments contain 5%Np in addition to 20%Pu-MOX. Other segments contain no Np. The objective of the irradiation tests is to obtain the restructuring data in the early beginning of life for SPF as well as power-to-melt test data for the potential study of SPF. At the same time introduction of modeling technique for irradiation performance analysis, fuel design, fuel fabrication is also important objective for JNC. Fabrication of irradiation test pins was completed till May 2003 in PSI. After transportation of the fuel pins to Petten, two times of irradiation were performed in January to March in 2004 and now post irradiation tests are in progress. Later two irradiations will be done till the autumn in 2004. This report summarized the basic plan, fuel design, and fabrication of irradiation test pins concerning this collaboration project. (author)

  11. Strategies and solutions in the temporary management of spent fuel in Spain

    International Nuclear Information System (INIS)

    Martinez Abad, J. E.; Rivera, M. I.

    2009-01-01

    The basic strategy for the spent fuel and HLW management contemplated in the Sixth General Radioactive Waste Plan focused on the centralised interim storage of spent fuel, based on proved dry storage system technologies, over the time periods required until their definitive or very long term management. Specially, the solution proposed as the most suitable for the Spanish case is the construction of a centralised interim spent fuel and HLW storage facility (ATC) for which as site is being searched. Until it becomes in operation, the interim spent fuel storage will be safety performed in the NPP reracked spent fuel pools or individual ISFSI constructed in the NPP site, in those cases additional storage capacity is required. (Author) 22 refs

  12. Development of TVSA VVER-1000 fuel

    International Nuclear Information System (INIS)

    Samoilov, O.; Kaydalov, V.; Romanov, A.; Falkov, A.; Morozkin, O.; Sholin, E.

    2013-01-01

    The TVSA fuel assemblies with a rigid angle-piece skeleton operate at 21 VVER-1000 units of Kalinin NPP, and Ukrainian, and Czech and Bulgarian NPPs. The total of more than 6,000 TVSA fuel assemblies have been fabricated. High lifetime performance has been achieved, namely, the maximum FA burnup is 65 MW∙day/kgU; maximum fuel rod burnup is 72 MW∙day/kgU; the lifetime is 50,000 EFPH. The TVSA fuel assembly is being improved to enhance its technical and economic performance and competitiveness of the Russian fuel for the VVER-1000 reactor: 1) Reliability and safety are being enhanced; repairability is being ensured. 2) High burnup levels in fuel are being ensured. 3) The uranium content in FAs is being increased. 4) The operational life is being extended. 5) Thermal-technical characteristics of FAs are being improved. The basic TVSA fuel assembly design evolved into the TVSA-PLUS with the fuel column elongated by 150 mm. The TVSA-PLUS fuel assembly has been in operation since 2010 at Kalinin NPP power units; an eighteen-month cycle is implemented at the uprated power of 104%. The TVSA-12PLUS fuel assembly has been developed with an elongated fuel column, optimized spacer grid positions (the spacer grid pitch is 340 mm) and with ensuring higher rigidity for the skeleton. It is provided for that fuel rods with the elevated uranium content and mixing intensifier grids will be used. The TVSA-T is developed for VVER-1000 reactor cores at the Temelin NPP. The TVSA-T is characterized by a load-carrying skeleton formed with angle-pieces and combined spacer grids that incorporate mixer grids. The TVSA-T design won the international tender to supply fuel to the Temelin NPP in the Czech Republic, and currently Temelin NPP Unit 1 and 2 are operating with the cores fully loaded with TVSA-Ts

  13. Reactor-specific spent fuel discharge projections: 1986 to 2020

    International Nuclear Information System (INIS)

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

    1987-03-01

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

  14. Nuclear safety in fuel-reprocessing plants

    International Nuclear Information System (INIS)

    Hennies, H.H.; Koerting, K.

    1976-01-01

    The danger potential of nuclear power and fuel reprocessing plants in normal operation is compared. It becomes obvious that there are no basic differences. The analysis of possible accidents - blow-up of an evaporator for highly active wastes, zircaloy burning, cooling failure in self-heating process solutions, burning of a charged solvent, criticality accidents - shows that they are kept under control by the plant layout. (HP) [de

  15. 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)

  16. Results on safety research for five years (from fiscal year 1996 to 2000). A field of nuclear fuel cycle

    International Nuclear Information System (INIS)

    2001-10-01

    This safety research carried out by the Japan Nuclear Cycle Development Institute (JNC) for five years ranged from fiscal year 1996 to 2000, was performed according to the safety research basic plan (from fiscal year 1996 to 2000) established on March, 1996 (revised again on May, 2000). This report was arranged on a field on nuclear fuel cycle (all subjects on fields of nuclear fuel facility, environmental radioactivity and radioactive wastes and a subject on nuclear fuel cycle in a field of seismic resistant and probabilistic safety assessment) by combining its research results for five years ranged from 1996 to 2000 fiscal year with general outlines on the safety research basic plan. Here were shown outlines on the safety research basic plan, aims and subjects on safety research at a field of nuclear fuel cycle, a list of survey sheets on safety research result, and survey sheets on safety research results. The survey sheets containing research field, title, organization, researcher name, researching period, names of cooperative organization, using facilities, research outline, research results, established contents, application, and research trends, are ranged to 21 items on nuclear fuel facility, 1 item on seismic resistance, 2 items on probabilistic safety assessment, 8 items on environmental radioactivity, and 20 items on radioactive wastes. (G.K.)

  17. Development of CANDU high-burnup fuel fabrication technology

    International Nuclear Information System (INIS)

    Sim, Ki Seob; Suk, H. C.; Kwon, H. I.; Ji, C. G.; Cho, M. S.; Chang, H. I.

    1997-07-01

    This study is focused on the achievement of the fabrication process improvement of CANFLEX-NU and for this purpose, following two areas of basic research were executed this year. 1) development of amorphous alloy for use in brazing of nuclear materials. 2) development of ECT techniques for the end-cap weld inspection. Also, preliminary feasibility analyses on the characteristics and handling techniques of CANFLEX-RU fuel were executed this year. - Selection of optimum conversion process of RU power -Characterization of the composition of RU power - Radiological characterization of RU power and sintered pellets - Compaction and sintering characteristics of RU power - Required special process for the production of CANFLEX-RU fuel - Development of technical specification for RU powder and pellets. In addition, technical support activities were performed for in-pile and out-pile fuel performance tests such as precision measurement of out-pile test fuel dimensions, establishment of quality control technique on fuel bundle by providing bundle kits to AECL for use in-pile irradiation tests in the NRU research reactor. (author). 57 refs., 16 tabs.,40 figs

  18. Photoelectrochemical solar water splitting: From basic principles to advanced devices

    Directory of Open Access Journals (Sweden)

    Bandar Y.Alfaifi

    2018-02-01

    Full Text Available Photoelectrochemical water splitting (PEC offers a promising path for sustainable generation of hydrogen fuel. However, improving solar fuel water splitting efficiency facing tremendous challenges, due to the energy loss related to fast recombination of the photogenerated charge carriers, electrode degradation, as well as limited light harvesting. This review focuses on the brief introduction of basic fundamental of PEC water splitting and the concept of various types of water splitting approaches. Numerous engineering strategies for the investgating of the higher efficiency of the PEC, including charge separation, light harvesting, and co-catalysts doping, have been discussed. Moreover, recent remarkable progress and developments for PEC water splitting with some promising materials are discussed. Recent advanced applications of PEC are also reviewed. Finally, the review concludes with a summary and future outlook of this hot field.

  19. Novel Blend Membranes Based on Acid-Base Interactions for Fuel Cells

    Directory of Open Access Journals (Sweden)

    Yongzhu Fu

    2012-10-01

    Full Text Available Fuel cells hold great promise for wide applications in portable, residential, and large-scale power supplies. For low temperature fuel cells, such as the proton exchange membrane fuel cells (PEMFCs and direct methanol fuel cells (DMFCs, proton-exchange membranes (PEMs are a key component determining the fuel cells performance. PEMs with high proton conductivity under anhydrous conditions can allow PEMFCs to be operated above 100 °C, enabling use of hydrogen fuels with high-CO contents and improving the electrocatalytic activity. PEMs with high proton conductivity and low methanol crossover are critical for lowering catalyst loadings at the cathode and improving the performance and long-term stability of DMFCs. This review provides a summary of a number of novel acid-base blend membranes consisting of an acidic polymer and a basic compound containing N-heterocycle groups, which are promising for PEMFCs and DMFCs.

  20. Basic strategical principles regarding Romanian power development after the year 2000

    International Nuclear Information System (INIS)

    Hotopeleanu, S.; Petrescu, C.

    1996-01-01

    This study deals with the the basic strategies for long term power planning in Romanian economy taking into account socio-economic factors, refurbishment and restructuring, having in view the requirements of free market economy transition. The main directions of this strategy refer to the development of the present power capacities based on fossil fuel burning and hydroelectric power and also to the development of nuclear power plants based on CANDU type reactors. (author) 3 tabs

  1. A code system to generate multigroup cross-sections using basic data

    International Nuclear Information System (INIS)

    Garg, S.B.; Kumar, Ashok

    1978-01-01

    For the neutronic studies of nuclear reactors, multigroup cross-sections derived from the basic energy point data are needed. In order to carry out the design based studies, these cross-sections should also incorporate the temperature and fuel concentration effects. To meet these requirements, a code system comprising of RESRES, UNRES, FIGERO, INSCAT, FUNMO, AVER1 and BGPONE codes has been adopted. The function of each of these codes is discussed. (author)

  2. Basic plans of nuclear energy development and utilization for fiscal 1982 (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Prime Minister, concerning the basic plans of nuclear energy development and utilization for fiscal 1982, was presented; the NSC has decided on the plans drawn up by the Prime Minister. Nuclear power generation as the nucleus of petroleum substitutes must be developed steadily. For the purpose, nuclear fuel cycle should be established, including the securing of uranium resources, uranium enrichment, fuel reprocessing, and waste management. The contents are as follows: the strengthening of nuclear safety measures, the promotion of nuclear power generation, the establishment of nuclear fuel cycle, the development of advanced types of reactors, the research on nuclear fusion, the research and development of nuclear powered ships, the promotion of radiation utilization, the strengthening of basis for nuclear energy development and utilization, the promotion of international cooperation, the strengthening of safeguard and nuclear material protection measures, fiscal 1982 budgets related to nuclear energy. (Mori, K.)

  3. Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2016-12-15

    The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

  4. Is fuel poverty in Ireland a distinct type of deprivation?

    OpenAIRE

    Watson, Dorothy; Maitre, Bertrand

    2014-01-01

    In this paper, we draw on the Central Statistics Office SILC data for Ireland to ask whether fuel poverty is a distinctive type of deprivation that warrants a fundamentally different policy response than poverty in general. We examine the overlap between fuel poverty (based on three self-report items) and poverty in general – with a particular emphasis on the national indicator of basic deprivation which is used in the measurement of poverty for policy purposes in Ireland. We examine changes ...

  5. Innovative microstructures in nuclear fuels

    International Nuclear Information System (INIS)

    Kutty, T.R.G.; Kumar, Arun; Kamath, H.S.

    2009-01-01

    For cleaner and safe nuclear power, new processes are required to design better nuclear fuels and make more efficient reactors to generate nuclear power. Therefore, one must understand how the microstructure changes during reactor operation. Accordingly, the materials scientists and engineers can then design and fabricate fuels with higher reliability and performance. Microstructure and its evolution are big unknowns in nuclear fuel. The basic requirements for the high performance of a fuel are: a) Soft pellets - To reduce Pellet clad mechanical interaction (PCMI) b) Large grain size - To reduce fission gas release (FGR). The strength of the pellet at room temperature is related to grain size by the Hall-Petch relation. Accordingly, the lower grain sized pellets will have high strength. But at high temperature (above equicohesive temperature) the grain boundaries becomes weaker than grain matrix. Since the small grain sized pellets have more grain boundary areas, these pellet become softer than pellet that have large grain sizes. Also as grain size decreases, creep rate of the fuel increases. Therefore, pellets with small grain size have higher creep rate and better plasticity. Therefore, these pellets will be useful to reduce the PCMI. On the other hand, pellet with large grain size is beneficial to reduce the fission gas release. In developing thermal reactor fuels for high burn-up, this factor should be taken into consideration. The question being asked is whether the microstructure can be tailored for irradiation hardening, fracture resistance, fission-gas release. This paper deals with the role played by microstructure for better irradiation performance. (author)

  6. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1982-06-01

    This publication is the third in a proposed series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. Subsequent issues in this series will update the information contained herein. This publication contains basically three kinds of information: (1) routes approved by the Commission for the shipment of irradiated reactor fuel, (2) information regarding any safeguards-significant incidents which have been reported to occur during shipments along such routes, and (3) cumulative amounts of material shipped

  7. Fabrication of Non-instrumented capsule for DUPIC simulated fuel irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.G.; Kang, Y.H.; Park, S.J.; Shin, Y.T. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    In order to develope DUPIC nuclear fuel, the irradiation test for simulated DUPIC fuel was planed using a non-instrumented capsule in HANARO. Because DUPIC fuel is highly radioactive material the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO was designed to remotely assemble and disassemble in hot cell. And then, according to the design requirements the non-instrumented DUPIC capsule was successfully manufactured. Also, the manufacturing technologies of the non-instrumented capsule for irradiating the nuclear fuel in HANARO were established, and the basic technology for the development of the instrumented capsule technology was accumulated. This report describes the manufacturing of the non-instrumented capsule for simulated DUPIC fuel. And, this report will be based to develope the instrumented capsule, which will be utilized to irradiate the nuclear fuel in HANARO. 26 refs., 4 figs. (Author)

  8. Partitioning of fissile and radio-toxic materials from spent nuclear fuel

    International Nuclear Information System (INIS)

    Bychkov, A.V.; Skiba, O.V.; Kormilitsyn, M.V.

    2007-01-01

    Full text of publication follows. The term ''partitioning'' means separation of one group of radwaste components from another. Such technological approaches are mainly applied to extraction of long-lived fission products (Tc, I) and minor actinides (Np, Am, Cm) from the waste arising from spent nuclear fuel reprocessing. Transmutation of the extracted minor actinides should be performed in a reactor or some accelerated systems. The combination of these technologies, partitioning and transmutation (P and T), will reduce the radiotoxicity of radwaste. In recent decades, partitioning has been directly linked to spent fuel reprocessing. Therefore, the basic investigations have been focused on the partitioning of liquid wastes arising from the PUREX process. These subjects have been the most developed ones, but the processes of fine aqueous separation generates an extra amount of liquid waste. This fact has an effect on the nuclear fuel cycle economy. Therefore, some other advanced compact methods have also been studied. These are dry methods involving molten chlorides and fluorides, the methods based on a supercritical movable phase, etc. The report provides a brief review of information on the basic partitioning process flow-sheets developed in France, Japan, Russia and other countries. Recent approaches to partitioning have been mostly directed towards radio-toxic hazard reduction and ecology. In the future, partitioning should be closely bound up with reprocessing and other spent nuclear fuel management processes. Reprocessing/partitioning should also be aimed at solving the problems of safety (non-proliferation) and economy in a closed fuel cycle. It is necessary to change a future ''technological philosophy'' of reprocessing and partitioning. The basic spent fuel components (U, Pu, Th) are to be extracted only for recycling in a closed nuclear fuel cycle. If these elements are regarded as a waste, additional expenses are required for transmutation. If we consider

  9. Neutron physics computation of CERCA fuel elements for Maria Reactor

    International Nuclear Information System (INIS)

    Andrzejewski, K.J.; Kulikowska, T.; Marcinkowska, Z.

    2008-01-01

    Neutron physics parameters of CERCA design fuel elements were calculated in the framework of the RERTR (Reduced Enrichment for Research and Test Reactors) program for Maria reactor. The analysis comprises burnup of experimental CERCA design fuel elements for 4 cycles in Maria Reactor To predict the behavior of the mixed core the differences between the CERCA fuel (485 g U-235 as U 3 Si 2 , 5 fuel tubes, low enrichment 19.75 % - LEU) and the presently used MR-6 fuel (430 g as UO 2 , 6 fuel tubes, high enrichment 36 % - HEU) had to be taken into account. The basic tool used in neutron-physics analysis of Maria reactor is program REBUS using in its dedicated libraries of effective microscopic cross sections. The cross sections were prepared using WIMS-ANL code, taking into account the actual structure, temperature and material composition of the fuel elements required preparation of new libraries.The problem is described in the first part of the present paper. In the second part the applicability of the new library is shown on the basis of the fuel core computational analysis. (author)

  10. Studies on the fission products behavior during dissolution process of BWR spent fuel

    International Nuclear Information System (INIS)

    Sato, K.; Nakai, E.; Kobayashi, Y.

    1987-01-01

    In order to obtain basic data on fission products behavior in connection with the head end process of fuel reprocessing, especially to obtain better understanding on undissolved residues, small scale dissolution studies were performed by using BWR spent fuel rods which were irradiated as monitoring fuel rods under the monitoring program for LWR fuel assembly performance entitled PROVING TEST ON RELIABILITY OF FUEL ASSEMBLY . The Zircaloy-2 claddings and the fuel pellets were subjected individually to the following studies on 1) release of fission products during dissolution process, 2) characterization of undissolved residues, and 3) analysis of the claddings. This paper presents comprehensive descriptions of the fission products behavior during dissolution process, based on detailed and through PIE conducted by JNFS under the sponsorship of MITI (Ministry of International Trade and Industry)

  11. Safety aspects of LWR fuel reprocessing and mixed oxide fuel fabrication plants

    International Nuclear Information System (INIS)

    Fischer, M.; Leichsenring, C.H.; Herrmann, G.W.; Schueller, W.; Hagenberg, W.; Stoll, W.

    1977-01-01

    The paper is focused on the safety and the control of the consequences of credible accidents in LWR fuel reprocessing plants and in mixed oxide fuel fabrication plants. Each of these plants serve for many power reactor (about 50.000 Mwel) thus the contribution to the overall risk of nuclear energy is correspondingly low. Because of basic functional differences between reprocessing plants, fuel fabrication plants and nuclear power reactors, the structure and safety systems of these plants are different in many respects. The most important differences that influence safety systems are: (1) Both fuel reprocessing and fabrication plants do not have the high system pressure that is associated with power reactors. (2) A considerable amount of the radioactivity of the fuel, which is in the form of short-lived radionuclides has decayed. Therefore, fuel reprocessing plants and mixed oxide fuel fabrication plants are designed with multiple confinement barriers for control of radioactive materials, but do not require the high-pressure containment systems that are used in LWR plants. The consequences of accidents which may lead to the dispersion of radioactive materials such as chemical explosions, nuclear excursions, fires and failure of cooling systems are considered. A reasonable high reliability of the multiple confinement approach can be assured by design. In fuel reprocessing plants, forced cooling is necessary only in systems where fission products are accumulated. However, the control of radioactive materials can be maintained during normal operation and during the above mentioned accidents, if the dissolver off-gas and vessel off-gas treatment systems provide for effective removal of radioactive iodine, radioactive particulates, nitrogen oxides, tritium and krypton 85. In addition, the following incidents in the dissolver off-gas system itself must be controlled: failures of iodine filters, hydrogen explosion in O 2 - and NOsub(x)-reduction component, decomposition of

  12. Licensing of spent fuel storage facility including its physical protection in the Czech Republic

    International Nuclear Information System (INIS)

    Fajman, V.; Sedlacek, J.

    1992-01-01

    The current spent fuel management policies as practised in the Czech Republic are described, and the conception of the fuel cycle back end is outlined. The general principles and the legislative framework are explained of the licensing process concerning spent fuel interim storage facilities, including the environmental impact assessment component. The history is outlined of the licensing process for the spent fuel storage facility at the Dukovany NPP site, including the licensing of the transport and storage cask. The basic requirements placed on the physical safeguarding of the facility and on the licensing process are given. (J.B.). 13 refs

  13. Contemporary strategy for external nuclear fuel cycle development: An analysis of the work of the IAEA NMFCTS

    International Nuclear Information System (INIS)

    Nechaev, A.F.

    1989-01-01

    The section's program includes four basic areas of activity: (1) nuclear fuel ore resources; (2) processing nuclear and reactor materials; (3) reactor fuel design, fabrication and behavior; and (4) spent nuclear fuel handling. The paper discusses the present-day condition and tendencies in the development of the nuclear fuel cycle and characteristics of international collaboration, including initial stages of the reactor fuel cycle, reactor fuel technology, and spent nuclear fuel handling. In recent years, the IAEA has made active efforts to improve international collaboration in accord with contemporary needs, and the purpose of this survey consists of showing a few concrete results achieved by the NMFCTS in this regard

  14. Back end of the fuel cycle

    International Nuclear Information System (INIS)

    Wolfe, B.; Lambert, R.W.

    1975-01-01

    At present, that portion of the nuclear fuel cycle involving reprocessing, waste management, and mixed-oxide fuel fabrication is in an unsettled state. Government regulatory requirements with respect to all aspects of the back end of the fuel cycle are still being formulated, and there is little positive experience on the operation of commercial reprocessing or mixed-oxide fabrication plants. In view of this unsettled situation, it will be difficult to meet the reprocessing and mixed-oxide fabrication needs of the next decade in the pattern previously anticipated. The costs in the back end of the fuel cycle are much higher than had been anticipated several years ago, a situation similar to that of almost all large endeavors in this country. On the other hand, the added costs are small relative to total power costs and do not affect the economic advantage of nuclear power as compared to other power sources. A rough economic analysis indicates that the question for the back end of the fuel cycle has changed from one of optimizing profitability to one of determining the most economic disposition of spent fuel. Long-term spent fuel storage is a practical and economically acceptable way to provide time for determining a sound course of action for the back end of the fuel cycle. Indeed, if one could count on a breeder economy before the end of the century, one possible course of action is to store light-water fuel until the plutonium can be used in breeders. However, for philosophical as well as practical reasons, it is important that the uncertainties in the course of action should be resolved as quickly as possible. Long-term storage should not be an excuse to delay resolution of the basic questions. (U.S.)

  15. Safety analysis report of uranium dioxide fuel laboratory, Nuclear Research Centre Inchas, Egypt

    International Nuclear Information System (INIS)

    Abdel-Azim, M.S.; Abdel-Halim, A.

    1987-07-01

    In the Nuclear Research Center Inchas a uranium dioxide fuel laboratory is planned and built by the AEA Cairo (Atomic Energy Authority). The layout of this fuel lab and the programmatical contents are subject to the bilaterial cooperation between Egypt and the Federal Republic of Germany. In this report the safety analysis as basic items for the approval procedure are started in detail. (orig.) [de

  16. Integrated data management system for radioactive waste and spent fuel in Korea

    International Nuclear Information System (INIS)

    Shin, Young Ho

    2001-03-01

    An integrated data management system for the safe management of radioactive waste and spent fuel in Korea is developed to collect basic information, provide the framework for national regulation, and improve national competition and efficiency in the management of radioactive waste and spent fuel. This system can also provide public access to information such as a statistical graphs and integrated data from various waste generators to meet increased public needs and interests. So through the system, the five principles (independence, openness, clearance, efficiency and reliance) of safety regulation can be realized, and public understanding and reliance on the safety of spent fuel and radioactive waste management can be promoted by providing reliable information, it can ensure an openness within the international nuclear community and efficiently support international agreements among contracting parties by operating safe and efficient management of spent fuel and radioactive waste (IAEA joint convention on the safety of spent fuel management and on the safety of radioactive waste management), the system can compensate for the imperfections in safe regulation of radioactive waste and spent fuel management related to waste generation, storage and disposal, and make it possible to holistic control and finally re-organize the basic framework of KINS's intermediate and long term research organization and trends, regarding waste management policy is to integrate safe management and unit safe disposal. For this objectives, benchmark study was performed on similar data base system worldwide and data specification with major input/output data during the first phase of this project

  17. Integrated data management system for radioactive waste and spent fuel in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Park, Yong Taek [Korea Power Engineering Co., Inc., Yongin (Korea, Republic of)

    2002-05-15

    An integrated data management system for the safe management of radioactive waste and spent fuel in Korea is developed to collect basic information, provide the framework for national regulation and improve national competition and efficiency in the management of radioactive waste and spent fuel. This system can also provide public access to information such as a statistical graphs and integrated data from various waste generators to meet increased public needs and interests. Through the system, the five principles(independence, openness, clearance, efficiency and reliance) of safety regulation can be realized and public understanding and reliance on the safety of spent fuel and radioactive waste management can be promoted. By providing reliable information and openness within the international nuclear community can be ensured and efficient support of international agreements among contracting parties can be ensured. By operating safe and efficient management of spent fuel and radioactive waste (IAEA joint convention on the safety of spent fuel management and on the safety of radioactive waste management), the system can compensate for the imperfections in safe regulation of radioactive waste and spent fuel management related to waste generation, storage and disposal, and make it possible for holistic control and reorganization of the basic framework of KINS's intermediate and long term research organization and trends, regarding waste management policy so as to integrate safe management and unit safe disposal. To meet this objectives, design of the database system structure and the study of input/output data validation and verification methodology was performed during the second phase of this project.

  18. Life-cycle of fuel peat

    International Nuclear Information System (INIS)

    Leijting, J.; Silvo, K.

    1998-01-01

    The share of peat in the primary energy supply in Finland in 1996 was about 6.5 % and the area used for peat production was about 535 km 2 , corresponding to about 0.5 % of the original peatland area of Finland. Fuel peat production is hence a significant form of using natural resources. About 1.4 % of the total peatland area has been reserved for peat production. Approximately 95 % of the peat excavated in Finland is used as fuel peat, and 5 % as horticultural peat. As raw material and fuel peat can be considered to be slowly renewable material. The environmental impacts of fuel peat production, transportation and peat combustion were evaluated in this research by methods used in life-cycle assessment. Preparation and production phases of peat production areas, fuel peat transportation to power plants, combustion of peat in power plants, and disposal of the ashes formed the basis for the investigation. Data collected in 1994-1996 was used as the basic material in the research. Special attention was paid to the estimation of greenhouse gas balance when using a virgin bog and the forest drained peatland areas as starting points. Post-production use of peatlands were not inspected in the life-cycle assessment. The work was carried out in 1997 in cooperation with Vapo Oy. The regional environmental centers, VTT and Helsinki and Joensuu Universities assisted significantly in acquisition of the material and planning of the work 3 refs

  19. System description of the Basic MRS System for the FY 1990 Systems Integration Program studies

    International Nuclear Information System (INIS)

    McKee, R.W.; Young, J.R.; Konzek, G.J.

    1991-07-01

    This document provides both functional and physical descriptions of a conceptual high-level waste management system defined as a Basic MRS System. Its purpose is to provide a basis for required system computer modeling and system studies initiated in FY 1990 under the Systems Integration Program of the Office of Civilian Radioactive Waste Management Office (OCRWM). Two specific systems studies initiated in FY 1990, the Reference System Performance Evaluation and the Aggregate Receipt Rate Study, utilize the information in this document. The Basic MRS System is the current OCRWM reference high-level radioactive wastes repository system concept. It is designed to accept 3000 MTU per year of spent fuel and 400 equivalent MTU per year of high-level wastes. The Basic MRS System includes a storage-only MRS that provides for a limited amount of commercial spent fuel storage capacity prior to acceptance by the geologic repository for disposal. This document contains both functional descriptions of the processes in the waste management system and physical descriptions of the equipment and facilities necessary for performance of those processes. The basic MRS system contains all system components, from the waste storage facilities of the waste generators to the underground facilities for final disposal of the wastes. The major facilities in the system are the waste generator waste storage facilities, an MRS facility that provides interim storage wastes accepted from the waste generators, a repository facility that packages the wastes and then emplaces them in the geologic repository, and the transportation equipment and facilities for transporting the waste between these major facilities

  20. Hibiscus fiber carbon for fuel cell device material

    International Nuclear Information System (INIS)

    Nanik Indayaningsih; Anne Zulfia; Dedi Priadi; Suprapedi

    2010-01-01

    The objective of this research is carbon of hibiscus fibers for the application as basic material of fuel cell device. The carbon is made using a pyrolysis process in inert gas (nitrogen) for 1 hour at temperature of 500 °C, 700 °C and 900 °C. The X-Ray Diffractometer (XRD), Scanning Electron Microscope (SEM) and Impedance-Capacitance-Resistance-meter are used to find out the microstructure, morphology and electrical properties respectively. The results of the experiment showed that the carbon had a structure of amorphous, and as the semiconductor material the electrical conductivity was 5 x 10"-"5 S.cm"-"1 to 4.9 x 10"-"5 S.cm"-"1 increasing in accordance with the pyrolysis temperature. The morphology resembled to plaited mats constructed by porous fibers having width of 50 µm to 300 µm, thickness of 25 µm to 35 µm, and the porous size of 0.5 µm to 5 µm. This morphology enables carbon to be applied as a candidate for a basic material of the Proton Exchange Membrane Fuel Cell. (author)

  1. Biodiesel Basics

    Energy Technology Data Exchange (ETDEWEB)

    Putzig, Mollie [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-10-04

    This fact sheet (updated for 2017) provides a brief introduction to biodiesel, including a discussion of biodiesel blends, which blends are best for which vehicles, where to buy biodiesel, how biodiesel compares to diesel fuel in terms of performance, the difference between biodiesel and renewable diesel, how biodiesel performs in cold weather, whether biodiesel use will plug vehicle filters, how long-term biodiesel use may affect engines, biodiesel fuel standards, and whether biodiesel burns cleaner than diesel fuel. The fact sheet also dismisses the use of vegetable oil as a motor fuel.

  2. Biodiesel Basics

    Energy Technology Data Exchange (ETDEWEB)

    None

    2017-09-01

    This fact sheet (updated for 2017) provides a brief introduction to biodiesel, including a discussion of biodiesel blends, which blends are best for which vehicles, where to buy biodiesel, how biodiesel compares to diesel fuel in terms of performance, the difference between biodiesel and renewable diesel, how biodiesel performs in cold weather, whether biodiesel use will plug vehicle filters, how long-term biodiesel use may affect engines, biodiesel fuel standards, and whether biodiesel burns cleaner than diesel fuel. The fact sheet also dismisses the use of vegetable oil as a motor fuel.

  3. Comparison of two thorium fuel cycles for use in light water prebreeder/breeder reactor systems (AWBA Development Program)

    International Nuclear Information System (INIS)

    Merriman, F.C.; McCoy, D.F.; Boyd, W.A.; Dwyer, J.R.

    1983-05-01

    Light water prebreeder/breeder conceptual reactor systems have been developed which have the potential to significantly improve the fuel utilization of present generation light water reactors. The purpose of this study is to describe and compare two possible types of thorium fuel cycles for use in these light water prebreeder and breeder concepts. The two types of thorium fuel cycles basically differ in the fuel rod design used in the prebreeder cores and the uranium isotopic concentration of fuel supplied to the breeder cores

  4. Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

    International Nuclear Information System (INIS)

    Glasstone, S.; Sesonske, A.

    1994-01-01

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials

  5. Nonproliferation and safeguard considerations: Pebble Bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, conpare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  6. Proposals of new basic concepts on safety and radioactive waste and of new High Temperature Gas-cooled Reactor based on these basic concepts

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Masuro, E-mail: ogawa.masuro@jaea.go.jp

    2016-11-15

    , which can satisfy the conditions required is proposed based on the new basic concepts on safety and radioactive waste. It is indicated that the New HTGR with about 4 MW/m{sup 3} average power density and an YSZ stabilized fuel can response to social requirements for safety and environmental conservability against radioactive wastes, industrial requirements for economy, uranium resource sustainability and application diversity, and national requirements for non-proliferation and environmental protection from emitting carbon dioxide.

  7. Proposals of new basic concepts on safety and radioactive waste and of new High Temperature Gas-cooled Reactor based on these basic concepts

    International Nuclear Information System (INIS)

    Ogawa, Masuro

    2016-01-01

    , which can satisfy the conditions required is proposed based on the new basic concepts on safety and radioactive waste. It is indicated that the New HTGR with about 4 MW/m"3 average power density and an YSZ stabilized fuel can response to social requirements for safety and environmental conservability against radioactive wastes, industrial requirements for economy, uranium resource sustainability and application diversity, and national requirements for non-proliferation and environmental protection from emitting carbon dioxide.

  8. Advanced nuclear fuel cycles - Main challenges and strategic choices

    International Nuclear Information System (INIS)

    Le Biez, V.; Machiels, A.; Sowder, A.

    2013-01-01

    A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness

  9. Advanced nuclear fuel cycles - Main challenges and strategic choices

    Energy Technology Data Exchange (ETDEWEB)

    Le Biez, V. [Corps des Mines, 35 bis rue Saint-Sabin, F-75011 Paris (France); Machiels, A.; Sowder, A. [Electric Power Research Institute, Inc. 3420, Hillview Avenue, Palo Alto, CA 94304 (United States)

    2013-07-01

    A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness.

  10. Actinide Sciences at ITN - Basic Studies in Chemistry with Potential Interest for Partitioning, Fuel Fabrication and More

    International Nuclear Information System (INIS)

    Almeida, M.; Dias, M.; Goncalves, A.P.; Henriques, M.S.; Lopes, E.B.; Pereira, L.C.J.; Santos, I.C.; Verbovytskyy, Y.; Waerenborgh, J.C.; Branco, J.B.; Carretas, J.M.; Cruz, A.; Ferreira, A.C.; Gasche, T.A.; Leal, J.P.; Lopes, G.; Lourenco, C.; Marcalo, J.; Maria, L.; Monteiro, B.; Mora, E.; Pereira, C.C.L.; Paiva, I.

    2010-01-01

    The current activities in the area of actinide chemistry at ITN, comprising basic research studies in inorganic and organometallic chemistry, catalysis, gas-phase ion chemistry, thermochemistry, and solid state chemistry, are briefly described. Actinide (and lanthanide) chemistry studies at ITN will be pursued connecting basic research with potential applications in nuclear and non-nuclear areas. (authors)

  11. At-reactor storage of spent fuel for life-of-plant

    International Nuclear Information System (INIS)

    Fuierer, A.A.

    1990-01-01

    The management of commercial spent fuel is a fairly broad topic beginning with the discharge from a reactor, its storage on-site, its transport from the reactor site to a U.S. Department of Energy (DOE) facility, and its ultimate disposal in a geologic repository. This paper discusses spent-fuel management in the at-reactor phase. There are two basic methods for at-reactor storage of spent fuel. The first is wet storage in a pool, and the second is dry storage external to the plant in some form of cask or vault. Spent-fuel consolidation will impact the utility and the DOE waste system. Some of these impacts have a positive effect and some have a negative effect, and each will vary somewhat for each utility. Spent-fuel consolidation and life-of-plant storage will be an increased burden to utilities but will likely result in significant cost savings to the overall waste management system and by proper integration can result in significant institutional benefits

  12. Nuclear fuel quality assurance

    International Nuclear Information System (INIS)

    1976-01-01

    Full text: Quality assurance is used extensively in the design, construction and operation of nuclear power plants. This methodology is applied to all activities affecting the quality of a nuclear power plant in order to obtain confidence that an item or a facility will perform satisfactorily in service. Although the achievement of quality is the responsibility of all parties participating in a nuclear power project, establishment and implementation of the quality assurance programme for the whole plant is a main responsibility of the plant owner. For the plant owner, the main concern is to achieve control over the quality of purchased products or services through contractual arrangements with the vendors. In the case of purchase of nuclear fuel, the application of quality assurance might be faced with several difficulties because of the lack of standardization in nuclear fuel and the proprietary information of the fuel manufacturers on fuel design specifications and fuel manufacturing procedures. The problems of quality assurance for purchase of nuclear fuel were discussed in detail during the seminar. Due to the lack of generally acceptable standards, the successful application of the quality assurance concept to the procurement of fuel depends on how much information can be provided by the fuel manufacturer to the utility which is purchasing fuel, and in what form and how early this information can be provided. The extent of information transfer is basically set out in the individual vendor-utility contracts, with some indirect influence from the requirements of regulatory bodies. Any conflict that exists appears to come from utilities which desire more extensive control over the product they are buying. There is a reluctance on the part of vendors to permit close insight of the purchasers into their design and manufacturing procedures, but there nevertheless seems to be an increasing trend towards release of more information to the purchasers. It appears that

  13. Study on the high-precision laser welding technology of nuclear fuel elements processing

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Soo Sung; Yang, M. S.; Kim, W. K.; Lee, D. Y

    2001-01-01

    The proper welding method for appendage of bearing pads and spacers of PHWR nuclear fuel elements is considered important in respect to the soundness of weldments and the improvement of the performance of nuclear fuels during the operation in reactor. The probability of welding defects of the appendage parts is mostly apt to occur and it is connected directly with the safty and life prediction of the nuclear reactor in operation. Recently there has been studied all over the world to develope welding technology by laser in nuclear fuel processing, and the appendage of bearing pads and spacers of PHWR nuclear fuel elements. Therefore, the purpose of this study is to investigate the characteristics of the laser welded specimens and make some samples for the appendage of bearing pads of PHWR nuclear fuel elements. This study will be also provide the basic data for the fabrications of the appendage of bearing pads and spacers. Especially the laser welding is supposed to be used in the practical application such as precise materials manufacturing fields. In this respect this technology is not only a basic advanced technology with wide applications but also likely to be used for the development of directly applicable technologies for industries, with high potential benefits derived in the view point of economy and industry.

  14. Study on the high-precision laser welding technology of nuclear fuel elements processing

    International Nuclear Information System (INIS)

    Kim, Soo Sung; Yang, M. S.; Kim, W. K.; Lee, D. Y.

    2001-01-01

    The proper welding method for appendage of bearing pads and spacers of PHWR nuclear fuel elements is considered important in respect to the soundness of weldments and the improvement of the performance of nuclear fuels during the operation in reactor. The probability of welding defects of the appendage parts is mostly apt to occur and it is connected directly with the safty and life prediction of the nuclear reactor in operation. Recently there has been studied all over the world to develope welding technology by laser in nuclear fuel processing, and the appendage of bearing pads and spacers of PHWR nuclear fuel elements. Therefore, the purpose of this study is to investigate the characteristics of the laser welded specimens and make some samples for the appendage of bearing pads of PHWR nuclear fuel elements. This study will be also provide the basic data for the fabrications of the appendage of bearing pads and spacers. Especially the laser welding is supposed to be used in the practical application such as precise materials manufacturing fields. In this respect this technology is not only a basic advanced technology with wide applications but also likely to be used for the development of directly applicable technologies for industries, with high potential benefits derived in the view point of economy and industry

  15. Vertical integration in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Mommsen, J.T.

    1977-01-01

    Vertical integration in the nuclear fuel cycle and its contribution to market power of integrated fuel suppliers were studied. The industry subdivision analyzed is the uranium raw materials sector. The hypotheses demonstrated are that (1) this sector of the industry is trending toward vertical integration between production of uranium raw materials and the manufacture of nuclear fuel elements, and (2) this vertical integration confers upon integrated firms a significant market advantage over non-integrated fuel manufacturers. Under microeconomic concepts the rationale for vertical integration is the pursuit of efficiency, and it is beneficial because it increases physical output and decreases price. The Market Advantage Model developed is an arithmetical statement of the relative market power (in terms of price) between non-integrated nuclear fuel manufacturers and integrated raw material/fuel suppliers, based on the concept of the ''squeeze.'' In operation, the model compares net profit and return on sales of nuclear fuel elements between the competitors, under different price and cost circumstances. The model shows that, if integrated and non-integrated competitors sell their final product at identical prices, the non-integrated manufacturer returns a net profit only 17% of the integrated firm. Also, the integrated supplier can price his product 35% below the non-integrated producer's price and still return the same net profit. Vertical integration confers a definite market advantage to the integrated supplier, and the basic source of that advantage is the cost-price differential of the raw material, uranium

  16. Vibration characteristics analysis for HANARO fuel assembly

    International Nuclear Information System (INIS)

    Ryu, Jeong Soo; Yoon, Doo Byung

    2001-06-01

    For investigating the vibration characteristics of HANARO fuel assembly, the finite element models of the in-air fuel assemblies and flow tubes were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes and the fuel assemblies were developed. Then, modal analysis of the developed models was carried out. The analysis results show that the fundamental vibration modes of the in-air 18-element and 36-element fuel assemblies are lateral bending modes and its corresponding natural frequencies are 26.4Hz and 27.7Hz, respectively. The fundamental natural frequency of the in-water 18-element and 36-element fuel assemblies were obtained as 16.1Hz and 16.5Hz. For the verification of the developed finite element models, modal analysis results were compared with those obtained from the modal test. These results demonstrate that the natural frequencies of lower order modes obtained from finite element analysis agree well with those of the modal test and the estimation of the hydrodynamic mass is appropriate. It is expected that the analysis results will be applied as a basic data for the operation and management of the HANARO. In addition, when it is necessary to improve the design of the fuel assembly, the developed finite element models will be utilized as a base model for the vibration characteristic analysis of the modified fuel assembly

  17. Validating Westinghouse atom 16 x 16 and 18 x 18 PWR fuel performance

    International Nuclear Information System (INIS)

    Andersson, S.; Gustafson, J.; Jourdain, P.; Lindstroem, L.; Hallstadius, L.; Hofling, C.G.

    2001-01-01

    Westinghouse Atom designs and fabricates PWR fuel for all major European fuel types: 17 x 17 standard (12 ft) and 17 x 17 XL (14 ft) for Westinghouse type PWRs, and 16 x 16 and 18 x 18 fuel for Siemens type PWRs. The W Atom PWR fuel designs are based on the extensive Westinghouse CE PWR fuel experience from combustion engineering type PWRs. The W atom designs utilise basic design features from the W CE fuel tradition, such as all-Zircaloy mid grids and the proven ( 6 rod years) Guardian TM debris catcher, which is integrated in the bottom Inconel grid. Several new features have been developed to meet with stringent European requirements originating from requirements on very high burnup, in combination with low-leakage core operating strategies and high coolant temperatures. The overall reliability of the Westinghouse Atom PWR fuel is very high; no fuel failure has been detected since 1997. (orig.)

  18. Fuel Chemistry Division: progress report for 1985

    International Nuclear Information System (INIS)

    1988-01-01

    Fuel Chemistry Division was formed in May 1985 to give a larger emphasis on the research and development in chemistry of the nuclear fuel cycle. The areas of research in Fuel Chemistry Division are fuel development and its chemical quality control, understanding of the fuel behaviour and post irradiation examinations, chemistry of reprocessing and waste management processes as also the basic aspects of actinide and relevant fission product elements. This report summarises the work by the staff of the Division during 1985 and also some work from the previous periods which was not reported in the progress reports of the Radiochemistry Division. The work related to the FBTR fuel was one of the highlights during this period. In the area of process chemistry useful work has been carried out for processing of plutonium bearing solutions. In the area of mass spectrometry, the determination of trace constituents by spark source mass spectrometry has been a major area of research. Significant progress has also been made in the use of alpha spectromet ry techniques for the determination of plutonium in dissolver solution and other samples. The technology of plutonium utilisation is quite complex and the Division would continue to look into the chemical aspects of this technology and provide the necessary base for future developments in this area. (author)

  19. Reprocessing technology for present water reactor fuels

    International Nuclear Information System (INIS)

    McMurray, P.R.

    1977-01-01

    The basic Purex solvent extraction technology developed and applied in the U.S. in the 1950's provides a well-demonstrated and efficient process for recovering uranium and plutonium for fuel recycle and separating the wastes for further treatment and packaging. The technologies for confinement of radioactive effluents have been developed but have had limited utilization in the processing of commercial light water reactor fuels. Technologies for solidification and packaging of radioactive wastes have not yet been demonstrated but significant experience has been gained in laboratory and engineering scale experiments with simulated commercial reprocessing wastes and intermediate level wastes. Commercial scale experience with combined operations of all the required processes and equipment are needed to demonstrate reliable reprocessing centers

  20. Fuel Cell Car Design Project for Freshman Engineering Courses

    Science.gov (United States)

    Duke, Steve R.; Davis, Virginia A.

    2014-01-01

    In the Samuel Ginn College of Engineering at Auburn University, we have integrated a semester long design project based on a toy fuel cell car into our freshman "Introduction to Chemical Engineering Class." The project provides the students a basic foundation in chemical reactions, energy, and dimensional analysis that facilitates…

  1. Spent fuel disposal: is the underground the sole solution?

    International Nuclear Information System (INIS)

    Nachmilner, L.

    1997-01-01

    The following 4 major approaches to spent fuel disposal are discussed: permanent storage in an underground repository, reprocessing, partitioning and transmutation, and accelerator driven transmutation. It is concluded that underground disposal will remain the basic option for the near future, although pursuing the other methods is certainly worth while. (P.A.)

  2. The Neutronic Effect of Zr-rod and Sm2O3-disk in the Model of Kartini Reactor Fuel Element

    International Nuclear Information System (INIS)

    Edi-Trijono-Budisantoso; Bambang-Sumarsono; Tri-Wulan-Tjiptono

    2000-01-01

    The model of Kartini fuel element has been constructed using a variationof materials in it fuels. For basic model, the construction of fuel assemblyis a rod type fuel element with SS-304 cladding and UZrH (8.5 % wt U in 20 %enrichment) as a fuel meat and graphite reflector at the both ends of fuelmeat. By using is basic construction, the material then varied in 4 models asthe following: 1) The model with the Zr-rod in 5.76 mm diameter, is insertedat the axis of fuel meat; 2) The model with Sm 2 O 3 -disk in 1.28 mmthickness, is inserted at the both ends of fuel meat; 3) The compound ofmodel 1 and model 2; 4) The basic construction of cell without Zr-rod andSm 2 O 3 -disk. The neutronic-behavior of each model then calculated usingWIMSD4 and the result then represented as a graphics form of the correlationof parameters as the following: 1) The correlation between model and itneutron multiplication factor; 2) The comparison of neutron flux distributionfor each model. By the graphic can be concluded, that Zr-rod in the fuelelement can reduce power peaking factor to 82 % with the effect of neutronmultiplication factor reduced to 98 %, while the insertion of Sm 2 O 3 -disk atthe both ends of fuel meat can make neutron multiplication factor reduce to97 % but unfortunately the neutron peaking factor increase to 101 %. Theinsertion of Zr-rod in the axis of fuel meat, increase the safety margin offuel utilization, while the insertion of Sm 2 O 3 -disk at the both ends offuel meat can make it as the burn-able poison that maintain the neutronmultiplication factor stable for a long time utilization. (author)

  3. Targeting of Asian Americans and Pacific Islanders by the tobacco industry: results from the Minnesota Tobacco Document Depository

    Science.gov (United States)

    Muggli, M; Pollay, R; Lew, R; Joseph, A

    2002-01-01

    Objective: The study objective was to review internal tobacco industry documents written between 1985 and 1995 regarding the Asian American and Pacific Islander (AAPI) population in the USA. These documents detail opportunities and barriers to promotion of tobacco products, as viewed by the tobacco industry and its market research firms. Data sources/methods: Researchers reviewed tobacco industry documents from the document depository in Minneapolis, Minnesota and the tobacco industry's website, The Tobacco Archive, in a systematic fashion. A combined technique was employed using title keywords, dates, and names to search the 4(b) index. Findings: A review of internal tobacco company documents reveal that during the late 1980s, the industry and its market research firms recognised the importance of the AAPI community as a potential business market. Documents describe the population growth in this community, the high prevalence of smoking in countries of origin, high purchasing power of AAPI immigrants, cultural predisposition to smoking, opportunities afforded by the high proportion of retail businesses under AAPI ownership, barriers to developing the AAPI market, comprehensive campaigns, and political and lobbying efforts. Comprehensive campaigns were designed to integrate promotion efforts in AAPI consumer, retail, and business communities. Conclusions: The documents show that the tobacco industry developed specific promotion strategies to target the AAPI population. Tobacco control initiatives in the AAPI group have been slower to develop than in other targeted ethnic groups, and may benefit by increased awareness of industry methods to promote tobacco use. PMID:12198269

  4. Spent fuel reprocessing system security engineering capability maturity model

    International Nuclear Information System (INIS)

    Liu Yachun; Zou Shuliang; Yang Xiaohua; Ouyang Zigen; Dai Jianyong

    2011-01-01

    In the field of nuclear safety, traditional work places extra emphasis on risk assessment related to technical skills, production operations, accident consequences through deterministic or probabilistic analysis, and on the basis of which risk management and control are implemented. However, high quality of product does not necessarily mean good safety quality, which implies a predictable degree of uniformity and dependability suited to the specific security needs. In this paper, we make use of the system security engineering - capability maturity model (SSE-CMM) in the field of spent fuel reprocessing, establish a spent fuel reprocessing systems security engineering capability maturity model (SFR-SSE-CMM). The base practices in the model are collected from the materials of the practice of the nuclear safety engineering, which represent the best security implementation activities, reflect the regular and basic work of the implementation of the security engineering in the spent fuel reprocessing plant, the general practices reveal the management, measurement and institutional characteristics of all process activities. The basic principles that should be followed in the course of implementation of safety engineering activities are indicated from 'what' and 'how' aspects. The model provides a standardized framework and evaluation system for the safety engineering of the spent fuel reprocessing system. As a supplement to traditional methods, this new assessment technique with property of repeatability and predictability with respect to cost, procedure and quality control, can make or improve the activities of security engineering to become a serial of mature, measurable and standard activities. (author)

  5. Department of Energy: some aspects of basic research in the chemical sciences

    International Nuclear Information System (INIS)

    1979-01-01

    The basic research needs pertinent to DOE's specific mission are identified in the fields of combustion science, coal chemistry, reprocessing of reactor fuel and the disposal of radioactive waste, and analytical chemistry. Aspects of these fields which do not need DOE support are also identified in some cases. In addition recommendations are made on review procedures and funding, use of DOE laboratories by university and other extramural chemists, isotope availability, and critically evaluated data

  6. Biodiesel Basics

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-07-01

    This fact sheet provides a brief introduction to biodiesel, including a discussion of biodiesel blends and specifications. It also covers how biodiesel compares to diesel fuel in terms of performance (including in cold weather) and whether there are adverse effects on engines or other systems. Finally, it discusses biodiesel fuel quality and standards, and compares biodiesel emissions to those of diesel fuel.

  7. Operational and environmental hazards of the quality degradation of fuels for power plants

    Energy Technology Data Exchange (ETDEWEB)

    Szabo, I.

    1987-01-01

    Higher sulphur and other minerals content in coal increases hazardous emission and, simultaneously, the caloric efficiency of the fuel decreases. The basic production control aspect of coal mines should be to produce coal with lesser ash content and higher calorific value rather than the higher quantity output of mines. For testing purposes, continuous ash content and calorific value monitors and displays should be installed at the power plants before fuel combustion.

  8. Decommissioning of nuclear fuel cycle facilities. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    The objective of this Safety Guide is to provide guidance to regulatory bodies and operating organizations on planning and provision for the safe management of the decommissioning of non-reactor nuclear fuel cycle facilities. While the basic safety considerations for the decommissioning of nuclear fuel cycle facilities are similar to those for nuclear power plants, there are important differences, notably in the design and operating parameters for the facilities, the type of radioactive material and the support systems available. It is the objective of this Safety Guide to provide guidance for the shutdown and eventual decommissioning of such facilities, their individual characteristics being taken into account

  9. Wrapping up the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Rueth, N.

    1976-01-01

    Reprocessing basically entails recovering uranium and plutonium from spent fuel for reuse in light water reactors (LWRs). The wastes resulting from this process are transformed to products suitable for disposal. These endeavors extend uranium supplies and also reduce the size and amount of nuclear waste that must be stored. Reprocessing, however, also ''unlocks'' the fuel rods that currently imprison radioactive substances. If great care is not taken, it could rip open a Pandora's box, exposing reprocessing plant workers, the general public, and the environment to deadly radioactive substances. While no commercial reprocessing plants are currently operating in the U.S., a scenario for such efforts has been mapped out. The first step is to chop the fuel elements into small pieces so that the fuel is no longer protected by its corrosion-resistant cladding. The fuel is then dissolved away from the cladding with nitric acid. An organic solvent extracts plutonium and uranium, and additional solvent extraction or ion exchange operations separate the two substances. Plutonium is converted to plutonium oxide; uranium 235 is converted to uranium oxide. They can then be combined to a make mixed oxide fuel, and formed into fuel elements for use in nuclear reactors. Various wastes with varied levels of radioactivity are generated during these operations. All demand attention. Radioactive gaseous waste most often is filtered before release through tall stacks. Metal solid waste--debris, fuel claddings, and hulls--may be compacted or cryogenically crushed and stored at specially designed storage sites. Contaminated combustibles, such as paper and resins, are incinerated and the ash is fixed and packaged for storage. The plans of Allied-General Nuclear Services (AGNS), which claims to have the closest thing in the United States to a ready reprocessor are described

  10. A collaboration on extended INPRO case study of the DUPIC fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. H.; Yang, M. S.; Ko, W. I. (and others)

    2007-05-15

    Since 1992, KAERI, AECL, United States Department of States(USDOS) and IAEA have performed the DUPIC fuel cycle development activities as an international cooperative research program, which has now been chosen as a target nuclear system for an INPRO case study. This study will focus on a further improvement and modification of the basic principles, user requirements and acceptance limits, which are defined in the IAEA-TECDOC-1434 for an evaluation of its proliferation-resistance through a proliferation-resistance assessment of the whole fuel cycle of DUPIC based on the INPRO methodology. In order to further develop an evaluation method for a proliferation-resistance based on the INPRO methodology, the basic principles, user requirements and acceptance limits of a proliferation-resistance was reviewed and quantified. Then the evaluation model (material flow, facility scale, reference fuel, etc.) of the DUPIC fuel cycle was developed and a proliferation-resistance assessment of the DUPIC fuel cycle including the PWR fuel cycle was performed by using the revised INPRO methodology in the area of a proliferation resistance. Also, the recommendations for a further improvement of INPRO methodology were suggested through examining the INPRO methodology for a proliferation resistance assessment. Through the proliferation resistance assessment of the whole fuel cycle of DUPIC including the PWR fuel cycle, the proliferation-resistance methodology was updated and re-established. And based on its experience, The research results can be used not only to evaluate and determine the future domestic proliferation-resistant fuel cycles which were derived from the GEN{sub I}V or INPRO programs but also to improve a system design to enhance its proliferation resistance. The present results will be utilized for the development of an INPRO User's Manual which is being developed as an important issue by IAEA. The credibility of the research results were ensured by the IAEA

  11. A collaboration on extended INPRO case study of the DUPIC fuel cycle

    International Nuclear Information System (INIS)

    Park, J. H.; Yang, M. S.; Ko, W. I.

    2007-05-01

    Since 1992, KAERI, AECL, United States Department of States(USDOS) and IAEA have performed the DUPIC fuel cycle development activities as an international cooperative research program, which has now been chosen as a target nuclear system for an INPRO case study. This study will focus on a further improvement and modification of the basic principles, user requirements and acceptance limits, which are defined in the IAEA-TECDOC-1434 for an evaluation of its proliferation-resistance through a proliferation-resistance assessment of the whole fuel cycle of DUPIC based on the INPRO methodology. In order to further develop an evaluation method for a proliferation-resistance based on the INPRO methodology, the basic principles, user requirements and acceptance limits of a proliferation-resistance was reviewed and quantified. Then the evaluation model (material flow, facility scale, reference fuel, etc.) of the DUPIC fuel cycle was developed and a proliferation-resistance assessment of the DUPIC fuel cycle including the PWR fuel cycle was performed by using the revised INPRO methodology in the area of a proliferation resistance. Also, the recommendations for a further improvement of INPRO methodology were suggested through examining the INPRO methodology for a proliferation resistance assessment. Through the proliferation resistance assessment of the whole fuel cycle of DUPIC including the PWR fuel cycle, the proliferation-resistance methodology was updated and re-established. And based on its experience, The research results can be used not only to evaluate and determine the future domestic proliferation-resistant fuel cycles which were derived from the GEN I V or INPRO programs but also to improve a system design to enhance its proliferation resistance. The present results will be utilized for the development of an INPRO User's Manual which is being developed as an important issue by IAEA. The credibility of the research results were ensured by the IAEA Consultant

  12. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    Henriquez, C; Navarro, G; Pereda, C

    2000-01-01

    Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

  13. Fission Product Release from Spent Nuclear Fuel During Melting

    International Nuclear Information System (INIS)

    Howell, J.P.; Zino, J.F.

    1998-09-01

    The Melt-Dilute process consolidates aluminum-clad spent nuclear fuel by melting the fuel assemblies and diluting the 235U content with depleted uranium to lower the enrichment. During the process, radioactive fission products whose boiling points are near the proposed 850 degrees C melting temperature can be released. This paper presents a review of fission product release data from uranium-aluminum alloy fuel developed from Severe Accident studies. In addition, scoping calculations using the ORIGEN-S computer code were made to estimate the radioactive inventories in typical research reactor fuel as a function of burnup, initial enrichment, and reactor operating history and shutdown time.Ten elements were identified from the inventory with boiling points below or near the 850 degrees C reference melting temperature. The isotopes 137Cs and 85Kr were considered most important. This review serves as basic data to the design and development of a furnace off-gas system for containment of the volatile species

  14. Calculation of Heat-Bearing Agent’s Steady Flow in Fuel Bundle

    Science.gov (United States)

    Amosova, E. V.; Guba, G. G.

    2017-11-01

    This paper introduces the result of studying the heat exchange in the fuel bundle of the nuclear reactor’s fuel magazine. The article considers the fuel bundle of the infinite number of fuel elements, fuel elements are considered in the checkerboard fashion (at the tops of a regular triangle a fuel element is a plain round rod. The inhomogeneity of volume energy release in the rod forms the inhomogeneity of temperature and velocity fields, and pressure. Computational methods for studying hydrodynamics in magazines and cores with rod-shape fuel elements are based on a significant simplification of the problem: using basic (averaged) equations, isobaric section hypothesis, porous body model, etc. This could be explained by the complexity of math description of the three-dimensional fluid flow in the multi-connected area with the transfer coefficient anisotropy, curved boundaries and technical computation difficulties. Thus, calculative studying suggests itself as promising and important. There was developed a method for calculating the heat-mass exchange processes of inter-channel fuel element motions, which allows considering the contribution of natural convection to the heat-mass exchange based on the Navier-Stokes equations and Boussinesq approximation.

  15. Analysis of burnt nuclear fuel elements by gamma-spectrometry

    International Nuclear Information System (INIS)

    Lammer, M.

    1978-01-01

    Gamma-spectrometry allows a non-destructive determination of the fission and activation product content of spent nuclear fuel. The concentration of some of these products depends significantly on the so-called fuel parameters which describe the irradiation history and the fuel composition. The use of these dependences for deriving ''unknown fuel parameters'' from measured fission product activities is investigated in this work. Relevant application fields are burnup determination, fuel testing and inspections within the nuclear materials safeguards programme. The present thesis investigates how these dependences can be used to derive unknown fuel parameters. The possibilities and basic limitations of deriving information from a measured gamma spectrum are considered on principle. The main conclusion is that only ratios of fission product activities allow the development of an interpretation method which is generally applicable to all types of fuel from different reactors. The dependence of activity ratios on cooling time, irradiation time, integrated and final neutron flux, fuel composition, as well as fission and breeding rates are then investigated and presented graphically in a way suitable for applicaton. These relationships can be used for the analysis of spent fuel, and the detailed procedures, which depend on the applicaton field, are worked out in this work. In order to test the interpretation methods, samples of nuclear fuel have been irradiated and the gamma spectra analysed. The methods developed in this work can be applied successfully to the analysis of burnt fuel in the frame of fuel testing programmes and to safeguards verification purposes. If however, apart from a gamma spectrum, no information on the investigated fuel is available, the above-mentioned parameters can be derived with low accuracy only. (author)

  16. Novel Fuel Cells for Coal Based Systems

    Energy Technology Data Exchange (ETDEWEB)

    Thomas Tao

    2011-12-31

    The goal of this project was to acquire experimental data required to assess the feasibility of a Direct Coal power plant based upon an Electrochemical Looping (ECL) of Liquid Tin Anode Solid Oxide Fuel Cell (LTA-SOFC). The objective of Phase 1 was to experimentally characterize the interaction between the tin anode, coal fuel and cell component electrolyte, the fate of coal contaminants in a molten tin reactor (via chemistry) and their impact upon the YSZ electrolyte (via electrochemistry). The results of this work will provided the basis for further study in Phase 2. The objective of Phase 2 was to extend the study of coal impurities impact on fuel cell components other than electrolyte, more specifically to the anode current collector which is made of an electrically conducting ceramic jacket and broad based coal tin reduction. This work provided a basic proof-of-concept feasibility demonstration of the direct coal concept.

  17. Rapsodie: A closed fuel cycle

    International Nuclear Information System (INIS)

    Levallet, E.H.; Costa, L.; Mougniot, J.C.; Robin, J.

    1977-01-01

    The Fortissimo Version of the core of the RAPSODIE fast reactor produces 40 MWTh. Since its start up in May 1970 in the CEN-CADARACHE its availability has stayed around 85%. Some of the mixed oxyde fuel pins UO 2 - 30% PuO 2 have already reached 150.000 MWd/t. The reprocessing is done in the pilot plant located in the La Hague Center and the plutonium obtained has already been re-used in the reactor. The Rapsodie-Fortissimo cycle is therefore now a closed cycle. This cycle is quite representative of fast reactor cycle characteristics and thus provides a remarkable research and development tool for the study of fabrication, in-reactor performances, transport, storage and reprocessing. These studies concern in particular the evolution of fission products and heavy isotopes content in fuel which controls both reprocessing schemes and intensity of emitted radiations. A program for the analysis of irradiated fuel has been developed either using samples collected all along the cycle, or following the actual reprocessing subassemblies. A set of basic data and calculation models has been established with two objectives: to give a better interpretation of the experimental program on one hand, and to extrapolate these results to the fuel cycle of fast reactors in general on the other hand. The first results have been quite encouraging up to now [fr

  18. Main concepts and objectives of fuel performance modelling and core development

    International Nuclear Information System (INIS)

    Lassmann, K.; O'Carroll, C.; Van de Laar, J.; Ray, I.

    1994-01-01

    The basic concepts of the fuel rod modelling was reviewed and their limitations are discussed.Three more general modelling concepts are briefly outlined: one-dimensional models versus 2-D or 3-D models; steady-state versus transient modelling; empirical versus mechanical models. The fuel rod behaviour is determined by thermal, mechanical and physical processes such as densification, swelling, gas release, irradiation damage etc. It is shown that the macroscopic behaviour of a fuel rod is to a large extent determined by the local UO 2 microstructure. Any fuel rod model must include the solution of the heat conduction equation and the principal mechanical equations i. e. equilibrium and compatibility, together with constitutive equations. The basic limitations from the different assumptions made for the solution of the governing equations are identified. It was shown that the thermal and the mechanical analysis are strongly coupled and therefore errors are propagated. The individual correlations and processes are also not error-free and in only some cases can individual errors be estimated. There is almost for all processes a good understanding of the dominating parameters but the unsolved problem is how local quantities such as pores and grain structure, stresses etc. evolve during the irradiation. Thus the wrong input data for local processes is considered as one of the main source of uncertainty. Fission gas release and swelling could not be described by the corresponding equations without introducing some free parameters. The center line predictions of two codes for FUMEX blind exercise are given. It is stated that since the D-COM exercise in 1984, fuel rod performance codes have been improved considerably and must be now considered as mature tools for further optimisation of fuel rods. 1 tab., 12 figs., 22 refs

  19. Main concepts and objectives of fuel performance modelling and core development

    Energy Technology Data Exchange (ETDEWEB)

    Lassmann, K; O` Carroll, C; Van de Laar, J; Ray, I [Commission of the European Communities, Karlsruhe (Germany). European Inst. for Transuranium Elements; Stefanova, S [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika; Chantoin, P [International Atomic Energy Agency, Vienna (Austria)

    1994-12-31

    The basic concepts of the fuel rod modelling was reviewed and their limitations are discussed.Three more general modelling concepts are briefly outlined: one-dimensional models versus 2-D or 3-D models; steady-state versus transient modelling; empirical versus mechanical models. The fuel rod behaviour is determined by thermal, mechanical and physical processes such as densification, swelling, gas release, irradiation damage etc. It is shown that the macroscopic behaviour of a fuel rod is to a large extent determined by the local UO{sub 2} microstructure. Any fuel rod model must include the solution of the heat conduction equation and the principal mechanical equations i. e. equilibrium and compatibility, together with constitutive equations. The basic limitations from the different assumptions made for the solution of the governing equations are identified. It was shown that the thermal and the mechanical analysis are strongly coupled and therefore errors are propagated. The individual correlations and processes are also not error-free and in only some cases can individual errors be estimated. There is almost for all processes a good understanding of the dominating parameters but the unsolved problem is how local quantities such as pores and grain structure, stresses etc. evolve during the irradiation. Thus the wrong input data for local processes is considered as one of the main source of uncertainty. Fission gas release and swelling could not be described by the corresponding equations without introducing some free parameters. The center line predictions of two codes for FUMEX blind exercise are given. It is stated that since the D-COM exercise in 1984, fuel rod performance codes have been improved considerably and must be now considered as mature tools for further optimisation of fuel rods. 1 tab., 12 figs., 22 refs.

  20. Some aspects of nuclear fuel economics

    International Nuclear Information System (INIS)

    Timm, M.

    1975-01-01

    The paper reviews the economic aspects of nuclear fuel based on the year 1975. The focus of attention is on the well-established light-water reactors, which for the time being, and also for the foreseeable future, account for the bulk of the world's nuclear power capacity. The author describes the principal economic aspects of the light-water fuel cycle components and discusses their future development to the extent that this can be foreseen. Various approaches to the formulation of the contract for the supply of light-water fuel are discussed on the basis of the established procedure. After some introductory comments on methods of calculating specific fuel costs, the author describes the main results of fuel cycle calculations. The economic aspects of nuclear fuel for other types of reactors are discussed in the sequence in which it is planned to introduce such reactors commercially. The economic data available apply to reactors ranging from the heavy-water-moderated type, through helium-cooled high-temperature reactors, to the fast breeders, and are based more and more on results of feasibility studies and less on practical operational experience. For this reason the author discusses only the basic data of importance, as viewed from the present-day stand-point, and mentions the emerging trends. It should be pointed out, however, that a comprehensive description of this kind can only show the situation at a given moment. Numerous political, economic and technological influences are in a state of permanent flux, hence economic data and developments may easily change within a short time. (author)

  1. Dry refabrication technology development of spent nuclear fuel

    International Nuclear Information System (INIS)

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

    2012-04-01

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

  2. Dry refabrication technology development of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-04-15

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

  3. Design of FCI Experiments to Understand Fuel Out-Pin Phenomena in the SFR

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Park, Seong Dae [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook; Bang, In Cheol [Chungang Univ., Seoul (Korea, Republic of)

    2014-05-15

    It is important to guarantee a passive nuclear safety regarding enhanced negative reactivity by fragmenting the molten fuel. In the SFR, it has a strong point that the negative reactivity is immediately introduced when the metal fuel is melted by the UTOP or ULOP accident. These characteristics of the metal fuel can prevent from progressing in severe accidents such as core disruptive accidents (CDA). As key phenomena in the accidents, fuel-coolant interaction (FCI) phenomena have been studied over the last few decades. Especially, several previous researches focused on instability and fragmentation of a core melt jet in water. However, the studies showed too limited phenomena to fully understand. In the domestic SFR technology development, researches for severe accidents tend to lag behind ones of other countries. Or, South Korea has a very basic level of the research such as literature survey. Recently, the SAS4A code, which was developed at Argonne National Laboratory (ANL) for thermal-hydraulic and neutronic analyses of power and flow transients in liquid-metal-cooled nuclear reactors (LMRs), is still under development to consider for a metal fuel. The other countries carried out basic experiments for molten fuel and coolant interactions. However, in a high temperature condition, methods for analysis of structural interaction between molten fuel and fuel cladding are very limited. The ultimate objective of the study is to evaluate the possibility of recriticality accident induced by fuel-coolant interaction in the SFR adopting metal fuel. It is a key point to analyze the molten-fuel behavior based on the experimental results which show fuel-coolant interaction with the simulant materials. It is necessary to establish the test facility, to build database, and to develop physical models to understand the FCI phenomena in the SFR; molten fuel-coolant interaction as soon as the molten fuel is ejected to the sodium coolant channel and molten fuel-coolant interaction

  4. Public information circular for shipments of irradiated reactor fuel. Revision 4

    International Nuclear Information System (INIS)

    1984-06-01

    This publication is the fifth in a series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. This publication contains basically three kinds of information: (1) routes recently approved (18 months) by the Commission for the shipment of irradiated reactor fuel; (2) information regarding any safeguards-significant incidents that may be (to date none have) reported during shipments along such routes; and (3) cumulative amounts of material shipped

  5. Thermodynamic aspects of the use of oxides as nuclear reactor fuels

    International Nuclear Information System (INIS)

    Potter, P.E.; Rand, M.H.

    1980-01-01

    The application of thermodynamics to help our understanding of the complex physical chemistry occurring in an irradiated nuclear fuel is reviewed. Emphasis is placed on current problems, under five topics: (1) Basic data: phase diagram and thermodynamic properties of PuO 2 and (U-Pu)O 2 ; (2) Mass transport: movement of Pu, U, O, C, and fission products; (3) Chemical effects of fission: reactions between fuel and clad; (4) Reactions with coolants; (5) Safety aspects: behvior above 3000 0 K. 46 references, 6 figures, 2 tables

  6. Recent advances on the production and utilization trends of bio-fuels: A global perspective

    International Nuclear Information System (INIS)

    Demirbas, M.F.; Balat, Mustafa

    2006-01-01

    Bio-fuels are important because they replace petroleum fuels. There are many benefits for the environment, economy and consumers in using bio-fuels. Bio-oil can be used as a substitute for fossil fuels to generate heat, power and/or chemicals. Upgrading of bio-oil to a transportation fuel is technically feasible, but needs further development. Bio-fuels are made from biomass through thermochemical processes such as pyrolysis, gasification, liquefaction and supercritical fluid extraction or biochemical. Biochemical conversion of biomass is completed through alcoholic fermentation to produce liquid fuels and anaerobic digestion or fermentation, resulting in biogas. In wood derived pyrolysis oil, specific oxygenated compounds are present in relatively large amounts. Basically, the recovery of pure compounds from the complex bio-oil is technically feasible but probably economically unattractive because of the high costs for recovery of the chemical and its low concentration in the oil

  7. Hygiene Basics

    Science.gov (United States)

    ... Staying Safe Videos for Educators Search English Español Hygiene Basics KidsHealth / For Teens / Hygiene Basics What's in this article? Oily Hair Sweat ... smell, anyway? Read below for information on some hygiene basics — and learn how to deal with greasy ...

  8. Concerning permission of change in nuclear fuel processing business of Japan Nuclear Fuel Co. , Ltd

    Energy Technology Data Exchange (ETDEWEB)

    1988-12-01

    In response to an inquiry on the title issue received on Jun. 17, 1988, the Nuclear Safety Commission made a study and submitted the findings to the Prime Minister on Jul. 21, 1988. The study was intended to determine the conformity of the permission to the applicable criteria specified in laws relating to control of nuclear material, nuclear fuel and nuclear reactor. The proposed modification plan included changes in the facilities in the No.1 processing building and changes in processing methods which were required to perform processing of blanket fuel assemblies for fast breeder reactor. It also included changes in the facilities in the No.2 building which were required to improve the processes. The safety study covered the anti-earthquake performance, fire/explosion prevention, criticality control, containment performance, radioactive waste disposal, and other major safety issues. Other investigations included exposure dose evaluation and accident analysis. Study results were examined on the basis of the Basic Guidelines for Nuclear Fuel Facilities Safety Review and the Uranium Processing Safety Review Guidelines. It was concluded that the modifications would not have adverse effect on the safety of the facilities. (Nogami, K.).

  9. Concerning permission of change in nuclear fuel processing business of Japan Nuclear Fuel Co., Ltd

    International Nuclear Information System (INIS)

    1988-01-01

    In response to an inquiry on the title issue received on Jun. 17, 1988, the Nuclear Safety Commission made a study and submitted the findings to the Prime Minister on Jul. 21, 1988. The study was intended to determine the conformity of the permission to the applicable criteria specified in laws relating to control of nuclear material, nuclear fuel and nuclear reactor. The proposed modification plan included changes in the facilities in the No.1 processing building and changes in processing methods which were required to perform processing of blanket fuel assemblies for fast breeder reactor. It also included changes in the facilities in the No.2 building which were required to improve the processes. The safety study covered the anti-earthquake performance, fire/explosion prevention, criticality control, containment performance, radioactive waste disposal, and other major safety issues. Other investigations included exposure dose evaluation and accident analysis. Study results were examined on the basis of the Basic Guidelines for Nuclear Fuel Facilities Safety Review and the Uranium Processing Safety Review Guidelines. It was concluded that the modifications would not have adverse effect on the safety of the facilities. (Nogami, K.)

  10. German Approach to Spent Fuel Management

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  11. Radial power density distribution of MOX fuel rods in the HBWR

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Joo, Hyung Kook; Lee, Byung Ho; Sohn, Dong Seong

    1999-07-01

    Two MOX fuel rods, which ar being fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with the Korea Atomic Energy Research Institute (KAERI), are going to be irradiated in the HBWR (Halden Boiling Water Reactor) from the beginning of 2000 in the framework of OECD Halden Reactor Programme (HRP) together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is a basic property in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR H BWR that calculates radial power density distribution for three MOX fuel rods have been developed subroutine FACTOR H BWR gives good agreement with the physics calculation except slight underprediction in the central part and a little overprediction at the outer part of the pellet. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. (author). 5 refs., 3 tabs., 24 figs

  12. Coal-fueled high-speed diesel engine development: Task 2, Market assessment and economic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kakwani, R. M.; Wilson, Jr., R. P.; Winsor, R. E.

    1991-12-01

    Based on the preliminary coal engine design developed, this task was conducted to identify the best opportunity(s) to enter the market with the future coal-fueled, high-speed diesel engine. The results of this market and economic feasibility assessment will be used to determine what specific heavy duty engine application(s) are most attractive for coal fuel, and also define basic economic targets for the engine to be competitive.

  13. Modeling the Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Jacobson, Jacob J.; Dunzik-Gougar, Mary Lou; Juchau, Christopher A.

    2010-01-01

    A review of existing nuclear fuel cycle systems analysis codes was performed to determine if any existing codes meet technical and functional requirements defined for a U.S. national program supporting the global and domestic assessment, development and deployment of nuclear energy systems. The program would be implemented using an interconnected architecture of different codes ranging from the fuel cycle analysis code, which is the subject of the review, to fundamental physical and mechanistic codes. Four main functions are defined for the code: (1) the ability to characterize and deploy individual fuel cycle facilities and reactors in a simulation, while discretely tracking material movements, (2) the capability to perform an uncertainty analysis for each element of the fuel cycle and an aggregate uncertainty analysis, (3) the inclusion of an optimization engine able to optimize simultaneously across multiple objective functions, and (4) open and accessible code software and documentation to aid in collaboration between multiple entities and facilitate software updates. Existing codes, categorized as annualized or discrete fuel tracking codes, were assessed according to the four functions and associated requirements. These codes were developed by various government, education and industrial entities to fulfill particular needs. In some cases, decisions were made during code development to limit the level of detail included in a code to ease its use or to focus on certain aspects of a fuel cycle to address specific questions. The review revealed that while no two of the codes are identical, they all perform many of the same basic functions. No code was able to perform defined function 2 or several requirements of functions 1 and 3. Based on this review, it was concluded that the functions and requirements will be met only with development of a new code, referred to as GENIUS.

  14. Development of the advanced CANDU technology -Development of CANDU advanced fuel fabrication technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Bum; Park, Choon Hoh; Park, Chul Joo; Kwon, Woo Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This project is carrying out jointly with AECL to develop CANFLEX fuel which can enhance reactor safety, fuel economy and can be used with various fuel cycles (natural U, slightly enriched U, other advanced fuel). The final goal of this research is to load the CANFLEX fuel in commercial CANDU reactor for demonstration irradiation. The annual portion of research activities performed during this year are followings ; The detail design of CANFLEX-NU fuel was determined. Based on this design, various fabrication drawings and process specifications were revised. The seventeen CANFLEX-NU fuel bundles for reactivity test in ZED-2 and out-pile test, two CANFLEX-SEU fuel bundles for demo-irradiation in NRU were fabricated. Advanced tack welding machine was designed and sequence control software of automatic assembly welder was developed. The basic researches related to fabrication processes, such as weld evaluation by ECT, effect of additives in UO{sub 2}, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author).

  15. Perspectives for practical application of the combined fuel kernels in VVER-type reactors

    International Nuclear Information System (INIS)

    Baranov, V.; Ternovykh, M.; Tikhomirov, G.; Khlunov, A.; Tenishev, A.; Kurina, I.

    2011-01-01

    The paper considers the main physical processes that take place in fuel kernels under real operation conditions of VVER-type reactors. Main attention is given to the effects induced by combinations of layers with different physical properties inside of fuel kernels on these physical processes. Basic neutron-physical characteristics were calculated for some combined fuel kernels in fuel rods of VVER-type reactors. There are many goals in development of the combined fuel kernels, and these goals define selecting the combinations and compositions of radial layers inside of the kernels. For example, the slower formation of the rim-layer on outer surface of the kernels made of enriched uranium dioxide can be achieved by introduction of inner layer made of natural or depleted uranium dioxide. Other potential goals (lower temperature in the kernel center, better conditions for burn-up of neutron poisons, better retention of toxic materials) could be reached by other combinations of fuel compositions in central and peripheral zones of the fuel kernels. Also, the paper presents the results obtained in experimental manufacturing of the combined fuel pellets. (authors)

  16. Reliability of the spent fuel identification for flask loading procedure used by COGEMA for fuel transport to La Hague

    International Nuclear Information System (INIS)

    Eid, M.; Zachar, M.; Pretesacque, P.

    1991-01-01

    The Spent Fuel Identification for Flask Loading (SFIFL) procedure designed by COGEMA is analysed and its reliability calculated. The reliability of the procedure is defined as the probability of transporting only approved fuel elements for a given number of shipments. The procedure describes a non-coherent system. A non-coherent system is the one in which two successive failures could result in a success, from the system mission point of view. A technique that describes the system with the help of its maximal cuts (states) is used for calculations. A maximal cut contains more than one failure which can split into two cuts (sub-states). Cuts splitting will enable us to analyse, in a systematic way, non-coherent systems with independent basic components. (author)

  17. Reliability of the spent fuel identification for flask loading procedure used by COGEMA for fuel transport to La Hague

    International Nuclear Information System (INIS)

    Eid, M.; Zachar, M.; Pretesacque, P.

    1990-01-01

    The Spent Fuel Identification for Flask Loading, SFIFL, procedure designed by COGEMA is analysed and its reliability is calculated. The reliability of the procedure is defined as the probability of transporting only approved fuel elements for a given number of shipments. The procedure describes a non-coherent system. A non-coherent system is the one in which two successive failures could result in a success, from the system mission point of view. A technique that describes the system with the help of its maximal cuts (states), is used for calculations. A maximal cut contains more than one failure can split into two cuts, (sub-states). Cuts splitting will enable us to analyse, in a systematic way, non-coherent systems with independent basic components. (author)

  18. Development of Dynamic Spent Nuclear Fuel Environmental Effect Analysis Model

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Ko, Won Il; Lee, Ho Hee; Cho, Dong Keun; Park, Chang Je

    2010-07-01

    The dynamic environmental effect evaluation model for spent nuclear fuel has been developed and incorporated into the system dynamic DANESS code. First, the spent nuclear fuel isotope decay model was modeled. Then, the environmental effects were modeled through short-term decay heat model, short-term radioactivity model, and long-term heat load model. By using the developed model, the Korean once-through nuclear fuel cycles was analyzed. The once-through fuel cycle analysis was modeled based on the Korean 'National Energy Basic Plan' up to 2030 and a postulated nuclear demand growth rate until 2150. From the once-through results, it is shown that the nuclear power demand would be ∼70 GWe and the total amount of the spent fuel accumulated by 2150 would be ∼168000 t. If the disposal starts from 2060, the short-term decay heat of Cs-137 and Sr-90 isotopes are W and 1.8x10 6 W in 2100. Also, the total long-term heat load in 2100 will be 4415 MW-y. From the calculation results, it was found that the developed model is very convenient and simple for evaluation of the environmental effect of the spent nuclear fuel

  19. Characteristics of fast reactor core designs and closed fuel cycle

    International Nuclear Information System (INIS)

    Poplavsky, V.M.; Eliseev, V.A.; Matveev, V.I.; Khomyakov, Y.S.; Tsyboulya, A.M.; Tsykunov, A.G.; Chebeskov, A.N.

    2007-01-01

    On the basis of the results of recent studies, preliminary basic requirements related to characteristics of fast reactor core and nuclear fuel cycle were elaborated. Decreasing reactivity margin due to approaching breeding ratio to 1, requirements to support non-proliferation of nuclear weapons, and requirements to decrease amount of radioactive waste are under consideration. Several designs of the BN-800 reactor core have been studied. In the case of MOX fuel it is possible to reach a breeding ratio about 1 due to the use of larger size of fuel elements with higher fuel density. Keeping low axial fertile blanket that would be reprocessed altogether with the core, it is possible to set up closed fuel cycle with the use of own produced plutonium only. Conceptual core designs of advanced commercial reactor BN-1800 with MOX and nitride fuel are also under consideration. It has been shown that it is expedient to use single enrichment fuel core design in this reactor in order to reach sufficient flattening and stability of power rating in the core. The main feature of fast reactor fuel cycle is a possibility to utilize plutonium and minor actinides which are the main contributors to the long-living radiotoxicity in irradiated nuclear fuel. The results of comparative analytical studies on the risk of plutonium proliferation in case of open and closed fuel cycle of nuclear power are also presented in the paper. (authors)

  20. Transition of Natural Frequencies of a Fuel Rod during Its Lifetime

    International Nuclear Information System (INIS)

    Kim, Hyeong Koo; Lee, Kyou Seok; Kim, Jeong Ha; Jeon, Sang Yoon

    2009-01-01

    The natural frequencies of a Pressurized Water Reactor (PWR) fuel rod are dependent on the geometrical and mechanical properties of fuel rod itself and its supporting conditions provided by spacer grids. By the way, these environmental parameters suffer remarkable change due to the plant operating conditions such as burnup, temperature, system pressure, and so on. It is inevitable, therefore, to be changed the natural frequencies of the fuel rod during its lifetime. In this paper, the transition of natural frequencies of the fuel rod for OPR1000 plants has been investigated considering fuel conditions associated with fuel life time. Basically for this investigation, three analysis models have been proposed representing beginning-of life (BOL) condition, middle-of-life (MOL) condition and end-of-life (EOL) condition including spacer grid supporting conditions. With these models, several modal analyses have been performed and the results have been compared with those of the test which has been carried out for verification of the analysis model. With these analyses and test, the changing vibration behavior of the PLUS7 fuel rod for OPR1000 during its life time has been discussed

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