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Sample records for cask analisis keselamatan

  1. KESELAMATAN PENGELASAN

    Directory of Open Access Journals (Sweden)

    Sarjito Jokosisworo

    2012-03-01

    Full Text Available Welding Inspector seringkali bekerja di tempat yang sama dengan tukang las mempunyai resiko terkena kecelakaan kerja seperti: terkena aliran listrik, jatuh, radiasi, mata terkena sinar ultra violet, dan partikel/debu di udara, asap dan gas serta kejatuhan benda-benda. Maka keselamatan kerja seorang welding inspector tidak boleh dipandang ringan. Welding Inspector dan pekerja lain yang bekerja di tempat pemotongan dan pengelasan harus memperhatikan semua faktor keselamatan, seperti penggunaan kacamata las, topi kerja/helm, pakaian kerja/wehrpak, sarung tangan kulit dan peralatan lain.

  2. Kajian Keselamatan Aktivitas Transportasi Laut terhadap Collision pada Bouy No. 15 Alur Pelayaran Barat Surabaya

    Directory of Open Access Journals (Sweden)

    Bimo Wira Para

    2015-03-01

    Full Text Available Kajian keselamatan aktivitas transportasi laut terhadap tubrukan kapal merupakan hal yang penting dilakukan bukan hanya untuk mengetahui safety level pada sebuah alur pelayaran, namun juga untuk mengurangi potensi kejadian tubrukan. Pelabuhan Tanjung Perak, Pelabuhan Gresik dan Pelabuhan Teluk Lamong, Jawa Timur, yang berada di Alur Pelayaran Barat Surabaya memiliki peranan yang besar dalam aktivitas ekspor impor dan perdagangan nasional jalur laut. Untuk lebih mengembangkan perekonomian nasional, Kementerian Perhubungan Republik Indonesia akan mengembangkan Rencana Induk Pelabuhan Tanjung Perak dan Sekitarnya Secara Terintegrasi, dimana salah satu pengembangannya adalah pembangunan dermaga yang akan dikelola oleh PT. Berlian Manyar Sejahtera, yang berada di sekitar Bouy No.15. Pada skripsi ini menyajikan kajian keselamatan terhadap tubrukan kapal di Bouy No.15 Alur Pelayaran Barat Surabaya dengan metode IWRAP dan  simulasi impak hasil tubrukan dilakukan dengan metode Finite Element Analysis. Penilaian ini bertujuan untuk mengetahui batas aman jumlah kapal yang diperbolehkan beroperasi di APBS setiap tahunnya. Data jumlah kapal yang berlayar di APBS didapatkan dari Pelindo sebagai Otoritas Pelabuhan di Indonesia. Hasil yang didapatkan berdasarkan analisis yang telah dilakukan terhadap head on collision, drifting collision, overtaking collision, dan crossing collision adalah sebesar 0.420, 0.940, 0.940, dan 0.605 secara berurutan, yang berarti frekuensi dari masing-masing jenis tubrukan dapat diterima jika mengacu pada keadaan future condition dimana frekuensi tubrukan dapat diterima bila bernilai dibawah satu. Dari analisis perhitungan yang telah dilakukan didapatkan kesimpulan bahwa jumlah maksimal kapal yang dapat berlayar di Alur Pelayaran Barat Surabaya pada future condition adalah sebanyak 49.640 kapal/tahun

  3. PENINGKATAN KINERJA SISTEM KESELAMATAN PASIF PADA REAKTOR NUKLIR DENGAN PENAMBAHAN KOMPONEN RVACS

    Directory of Open Access Journals (Sweden)

    A. G. Abdullah

    2014-07-01

    Full Text Available Kelengkapan sistem keselamatan pasif dan inheren pada reaktor lanjut merupakan prasyarat utama. Makalah ini mengeksplorasi hasil desain konseptual sistem pembuang sisa panas pada pusat listrik tenaga nuklir berjenis Very High-Temperature Reactor. Tujuan riset ini untuk merancang sistem pembuang sisa panas pusat listrik tenaga nuklir yang terdapat pada dinding reaktor. Studi kinerja Reactor Vessel Auxliary Cooling System (RVACS dilakukan pada dua jenis pendingin yaitu Timbal-Bismut dan Liquid Salt. Panas dari dinding reaktor dihapus melalui sirkulasi alamiah pada keadaan tunak. Analisis melibatkan sistem perpindahan panas secara radiasi, konduksi dan konveksi alami. Perhitungan perpindahan panas dilakukan pada elemen reaktor vessel, dinding luar guard vessel, dan pelat pemisah. Hasil analisis kecelakaan menunjukkan kedua jenis sistem pendingin reaktor dan sistem pasif sisa pembuangan panas cukup menghapus sisa panas hasil peluruhan dengan sirkulasi alami.ABSTRACTCompleteness of passive safety systems and inherent in advanced reactors is a major prerequisite. This paper explores the results of a conceptual design of the heat removal system at the nuclear power plant (NPP type Very High-Temperature Reactor. The purpose of this research was to design the reactor vessel auxiliary cooling system (RVACS of NPP located within the reactor walls. The RVACS performance study was conducted on two types of coolant: Lead-Bismuth and Liquid Salt. Heat was removed from the reactor vessel through the natural circulation in the steady state. Analyses of heat transfer systems involved radiation, conduction and natural convection. Heat transfer calculations were performed on the reactor vessel, guard vessel, and perforated plate. The results from the accident analysis showed that both types, the reactor coolant system and the passive residual heat removal system, adequately remove remaining heat of the decay by a natural circulation.

  4. Evaluasi dan Perancangan Sistem Manajemen Keselamatan dan Kesehatan Kerja (SMK3 Dalam Rangka Perbaikan Safety Behaviour Pekerja (Studi Kasus : PT. X, Sidoarjo

    Directory of Open Access Journals (Sweden)

    Dhinar Tiara Luckyta

    2012-09-01

    Full Text Available Kecelakaan kerja merupakan salah satu permasalahan yang melekat dalam dunia industri. Di Indonesia penyebab tingginya angka kecelakaan kerja salah satunya disebabkan karena kurangnya penerapan Sistem Manajemen Keselamatan dan Kesehatan Kerja (SMK3. PT. X, merupakan perusahaan yang belum menerapkan SMK3 sedangkan lingkungan kerja perusahaan cukup mengandung potensi bahaya. Kecelakaan kerja sering terjadi dikarenakan pekerja sering melakukan tindakan tidak aman (unsafe behaviour. Penelitian ini dilakukan untuk mengevaluasi SMK3 perusahaan dan mendapatkan penyebab dari unsafe behaviour pekerja dengan menggunakan root cause analysis serta solusi perbaikan digunakan HFMEA. Dari hasil analisis menunjukkan bahwa sebanyak 46 kriteria SMK3 belum terpenuhi. Sedangkan penyebab dari unsafe behaviour pekerja adalah fasilitas dan APD yang tidak nyaman untuk digunakan, suhu ruangan yang panas, kurangnya safety sign, kurangnya fungsi kontrol manajemen, dan tidak adanya peraturan yang tegas. Selain itu penelitian ini juga menghasilkan rancangan dan prosedur SMK3 yang mengacu pada Permenaker 05/MEN/1996

  5. Testing of Metal Cask and Concrete Cask

    International Nuclear Information System (INIS)

    In Japan, the first interim spent fuel storage facility (ISF) outside of nuclear power plant site in use of dual-purpose metal cask is being planned to start its commercial operation in 2012 in Mutsu city, Aomori prefecture. The CRIEPI (Central Research Institute of Electric Power Industry) has executed several study programs on demonstrative testing for interim storage of spent fuel, mainly related to metal cask and concrete cask storage technology to reflect in Japanese safety requirements for dry casks issued by NISA/METI (Nuclear and Industrial Safety Agency, Ministry of Economy and Trade Industry). On top of that, the Japan Nuclear Energy Safety Organization (JNES) has executed study programs on spent fuel integrity, etc. This paper introduces the summary of these research programs. (author)

  6. A cask fleet operations study

    Energy Technology Data Exchange (ETDEWEB)

    1988-03-01

    This document describes the cask fleet currently available to transport spent nuclear fuels. The report describes the proposed operational procedures for these casks and the vehicles intended to transport them. Included are techniques for loading the cask, lifting it onto the transport vehicle, preparing the invoices, and unloading the cask at the destination. The document concludes with a discussion on the maintenance and repair of the casks. (tem) 29 figs.

  7. Cask development, testing, and licensing

    International Nuclear Information System (INIS)

    The NuPac 125-B Rail Cask was developed to provide a safe means of transporting the damaged core of Three Mile Island Unit 2 from the TMI site at Middletown, PA, to the Idaho National Engineering laboratory (INEL) at Idaho Falls, ID. The development of the NuPac 125-B Rail Cask posed two engineering and technical management challenges; Licensing Strategy - The NuPac 125-B Rail Cask represented the first irradiated fuel rail cask developed within the United States in the past decade, a decade characterized by changing nuclear regulations, and Accelerated Schedule - The TMI-2 defueling schedule demanded a cask development schedule one-third as long as normally required. These challenges governed the overall development and licensing process for the cask. First, a high degree of conservation was incorporated into the design to allow quick, simplified demonstrations of adequacy to regulatory staff. Second, redundant design techniques were employed in all areas of uncertainty. The testing program eliminated performance uncertainties and validated predictions and predictive models. Drop tests of a quarter-scale model of the cask were conducted, and results were correlated with analytic predictions to verify structural and mechanical performance of the cask. Full-scale tests of the canisters were conducted to verify structural behavior of canister internals which provide criticality control. This paper describes the testing program for the NuPac 125-B Rail Cask, presents results therefrom, and correlates findings with Regulation 10 CFR 71 of the U.S. Nuclear Regulatory Commission

  8. Analisis Pengaruh Penerapan Sistem Manajemen Keselamatan dan Kesehatan Kerja OHSAS 18001 Terhadap Produktivitas Kerja Karyawan PT. Coca Cola Amatil Medan

    OpenAIRE

    Sitorus, Ummi Salamah

    2016-01-01

    Basically to doing his job, employees are not immune from the called risk in the work. Both the level of risk in mild or severe. To minimize the risk of loss is required a management system for managing safety and health of employees in order to increase employee productivity so companies can compete with other companies. In this case PT. Coca Cola Amatil Medan implement a safety and health Management System (SMK3) OHSAS 18001. The method used is quantitative research with assosiative app...

  9. Analisi matematica

    CERN Document Server

    Canuto, Claudio

    2014-01-01

    Il presente testo intende essere di supporto ad un secondo insegnamento di Analisi Matematica in quei corsi di studio (quali ad esempio Ingegneria, Informatica, Fisica) in cui lo strumento matematico parte significativa della formazione dell'allievo. I concetti e i metodi fondamentali del calcolo differenziale ed integrale in più variabili, le serie di funzioni e le equazioni differenziali ordinarie sono presentati con l'obiettivo primario di addestrare lo studente ad un loro uso operativo, ma critico. L'impostazione didattica dell'opera ricalca quella usata nel testo parallelo di Analisi Matematica I. La modalità di presentazione degli argomenti ne permette un uso flessibile e modulare. Lo stile adottato privilegia la chiarezza e la linearità dell'esposizione. Il testo organizzato su due livelli di lettura. Uno, più essenziale, permette allo studente di cogliere i concetti indispensabili della materia, di familiarizzarsi con le relative tecniche di calcolo e di trovare le giustificazioni dei principali r...

  10. CARRIER/CASK HANDLING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    E.F. Loros

    2000-06-23

    The Carrier/Cask Handling System receives casks on railcars and legal-weight trucks (LWTs) (transporters) that transport loaded casks and empty overpacks to the Monitored Geologic Repository (MGR) from the Carrier/Cask Transport System. Casks that come to the MGR on heavy-haul trucks (HHTs) are transferred onto railcars before being brought into the Carrier/Cask Handling System. The system is the interfacing system between the railcars and LWTs and the Assembly Transfer System (ATS) and Canister Transfer System (CTS). The Carrier/Cask Handling System removes loaded casks from the cask transporters and transfers the casks to a transfer cart for either the ATS or CTS, as appropriate, based on cask contents. The Carrier/Cask Handling System receives the returned empty casks from the ATS and CTS and mounts the casks back onto the transporters for reshipment. If necessary, the Carrier/Cask Handling System can also mount loaded casks back onto the transporters and remove empty casks from the transporters. The Carrier/Cask Handling System receives overpacks from the ATS loaded with canisters that have been cut open and emptied and mounts the overpacks back onto the transporters for disposal. If necessary, the Carrier/Cask Handling System can also mount empty overpacks back onto the transporters and remove loaded overpacks from them. The Carrier/Cask Handling System is located within the Carrier Bay of the Waste Handling Building System. The system consists of cranes, hoists, manipulators, and supporting equipment. The Carrier/Cask Handling System is designed with the tooling and fixtures necessary for handling a variety of casks. The Carrier/Cask Handling System performance and reliability are sufficient to support the shipping and emplacement schedules for the MGR. The Carrier/Cask Handling System interfaces with the Carrier/Cask Transport System, ATS, and CTS as noted above. The Carrier/Cask Handling System interfaces with the Waste Handling Building System for building

  11. PENERAPAN KESEHATAN DAN KESELAMATAN KERJA DI PUSKESMAS DI TIGA PROVINSI DI INDONESIA

    Directory of Open Access Journals (Sweden)

    Lusianawaty Tana

    2013-11-01

    kerja. Penelitian bertujuan mengevaluasi penerapan kesehatan dan keselamatan kerja (K3 dalam pencegahan penularan Mycobaterium tuberculosis di puskesmas dan hambatannya. Disain cross sectional, pada 50 puskesmas (PRM/PPM puskesmas rujukan mikroskopis/puskesmas pelaksana mandiri di provinsi Banten, Gorontalo, dan Kalimantan Selatan,  tahun 2012, data dikumpulkan melalui wawancara dan pengamatan. Data yang dikumpulkan meliputi karakteristik puskesmas, penerapan K3, kelengkapan pedoman, sarana dan prasarana. Penerapan K3 dalam rangka pencegahan TB paru di puskesmas belum seluruhnya sesuai dengan Pedoman Pencegahan Penularan M. tuberculosis WHO. Pelatihan pekerja puskesmas terhadap pencegahan penularan TB telah dilaksanakan oleh puskesmas. Penerapan K3 yang masih kurang adalah pelaksanaan kegiatan yang perlu dilakukan dan pemeriksaan kesehatan berkala/skrining TB masing-masing pada 58 % dan 56 % puskesmas. Alat promosi kesehatan terkait K3 masih kurang pada 26 % puskesmas.  Alat pelindung diri berupa sarung tangan dan masker tersedia pada hampir semua puskesmas  98% dan 96%. Sarana prasarana masih kurang pada 68 % puskesmas dan sarana prasarana laboratorium masih kurang pada 40 % puskesmas (n=50. Penerapan K3 di PRM/PPM perlu ditingkatkan dengan melengkapi sarana dan prasarana puskesmas dan laboratorium, serta alat promosi kesehatan.Kata kunci : puskesmas, kesehatan dan keselamatan kerja

  12. Cask fleet operations study

    International Nuclear Information System (INIS)

    The Nuclear Waste Policy Act of 1982 assigned to the Department of Energy's (DOE) Office of Civilian Waste Management the responsibility for disposing of high-level waste and spent fuel. A significant part of that responsibility involves transporting nuclear waste materials within the federal waste management system; that is, from the waste generator to the repository. The lead responsibility for transportation operations has been assigned to Oak Ridge Operations, with Oak Ridge National Laboratory (ORNL) providing technical support through the Transportation Operations Support Task Group. One of the ORNL support activities involves assessing what facilities, equipment and services are required to assure that an acceptable, cost-effective and safe transportation operations system can be designed, operated and maintained. This study reviews, surveys and assesses the experience of Nuclear Assurance Corporation (NAC) in operating a fleet of spent-fuel shipping casks to aid in developing the spent-fuel transportation system

  13. Cask fleet operations study

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    The Nuclear Waste Policy Act of 1982 assigned to the Department of Energy's (DOE) Office of Civilian Waste Management the responsibility for disposing of high-level waste and spent fuel. A significant part of that responsibility involves transporting nuclear waste materials within the federal waste management system; that is, from the waste generator to the repository. The lead responsibility for transportation operations has been assigned to Oak Ridge Operations, with Oak Ridge National Laboratory (ORNL) providing technical support through the Transportation Operations Support Task Group. One of the ORNL support activities involves assessing what facilities, equipment and services are required to assure that an acceptable, cost-effective and safe transportation operations system can be designed, operated and maintained. This study reviews, surveys and assesses the experience of Nuclear Assurance Corporation (NAC) in operating a fleet of spent-fuel shipping casks to aid in developing the spent-fuel transportation system.

  14. Design and operational experience of dry cask storage systems

    International Nuclear Information System (INIS)

    This paper (Power Point presentation) describes cask storage design features and available dry cask storage technology, cask types used for dry storage, design characteristics of CASTOR casks, the German licensing basis for cask storage systems, shielding requirements, thermal layout, mechanical design, criticality safety and containment, licensing procedure, operational experience of dry cask storage in Germany and worldwide

  15. ANALISIS KECELAKAAN REAKTOR AKIBAT KEGAGALAN SISTEM PEMBUANG PANAS PADA REAKTOR NUKLIR GENERASI IV

    Directory of Open Access Journals (Sweden)

    A. G. Abdullah

    2012-05-01

    Full Text Available Salah satu aspek terpenting dalam proses desain reaktor nuklir adalah aspek keselamatan reaktor. Sebelum membangun reaktor secara fisik, terlebih dahulu dibuat perencaaan perhitungan yang matang termasuk melakukan simulasi kinerja keselamatannya dalam menghadapi kemungkinan kecelakaan. Penelitian ini bertujuan untuk mengembangkan model simulasi kecelakaan Pembangkit Listrik Tenaga Nuklir (PLTN yang disebabkan  gagalnya sistem pembuang panas. Kecelakaan akibat gagalnya sistem pembuang panas dipicu oleh hilangnya kemampuan pendinginan dari pembangkit uap. Urutan kecelakaan ini diawali dengan hilangnya kemampuan reaktor untuk membuang panas dari loop pendingin sekunder. Selama kecelakaan, laju pembuangan panas mengalami penurunan sedangkan temperatur masukan pendingin mengalami peningkatan. Hasil simulasi memberikan gambaran bahwa reaktor dapat bertahan dari kecelakaan. Hasil analisis kecelakaan menunjukkan bahwa temperatur maksimum bahan bakar, selongsong dan pendingin memiliki batas keselamatan yang sangat besar.One of the most important aspects in nuclear reactor design process is the safety aspect. Advanced and accurate safety simulation must be performed before it can be built.  This research aims to develop a simulation model of Nuclear Power Plant (NPP accidents due to the loss of  heat sink system. Loss of heat sink accident was triggered by the loss of cooling capability of steam generators.  This  accident  sequence  began with the loss of the reactor’s ability to remove heat from the secondary cooling loop. During the accident, the heat dissipation rate decreased whereas the coolant inlet temperatures increased till a new equilibrium level. The analysis results of the accident showed that there are large safety margin to the maximum temperature of the fuel, cladding, and coolant.

  16. 78 FR 78693 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-12-27

    ... RIN 3150-AJ10 List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS Cask... final rule amended the NRC's spent fuel storage regulations by revising the Transnuclear, Inc. Standardized NUHOMS Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to...

  17. Analisis Statistika dengan SPSS

    OpenAIRE

    Mustari, Kahar

    2012-01-01

    Buku ini membahas tentang berbagai model analisis statistika yang dapat digunakan dalam berbagai bidang penelitian. Pembahasan buku ini dilengkapi pula dengan penerapan analisis statistika yang menggunakan program SPSS versi 20 sehingga memudahkan pembaca untuk menerapkannya. Oleh karena itu, buku ini penting dibaca oleh mahasiswa atau peneliti yang melakukan pengolahan data penelitian.

  18. GNS spent fuel cask experience

    International Nuclear Information System (INIS)

    The Gesellschaft fuer Nuklear-Service mbH (GNS), which is owned by German utilities, is responsible for the management of spent fuel and nuclear waste on behalf of the German utilities operating nuclear power plants. This paper describes the spent reactor fuel and waste shipping and/or storage casks that GNS manufacturers for nuclear facilities in Germany, and worldwide. So far more than 30 different casks have been produced in quantities ranging from one to several hundred of each type. GNS participates in the German Support Program to assist the International Atomic Energy Agency (IAEA) in developing verification procedures for dry storage casks containing spent fuel. This activity is also summarized

  19. GNS spent fuel cask experience

    Energy Technology Data Exchange (ETDEWEB)

    Weh, R. (Gesellschaft fuer Nuklear-Service mbH, Hannover (Germany))

    1993-05-01

    The Gesellschaft fuer Nuklear-Service mbH (GNS), which is owned by German utilities, is responsible for the management of spent fuel and nuclear waste on behalf of the German utilities operating nuclear power plants. This paper describes the spent reactor fuel and waste shipping and/or storage casks that GNS manufacturers for nuclear facilities in Germany, and worldwide. So far more than 30 different casks have been produced in quantities ranging from one to several hundred of each type. GNS participates in the German Support Program to assist the International Atomic Energy Agency (IAEA) in developing verification procedures for dry storage casks containing spent fuel. This activity is also summarized.

  20. SNF shipping cask shielding analysis

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.O.; Pace, J.V. III

    1996-01-01

    The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan.

  1. Initiatives in transport cask license

    Energy Technology Data Exchange (ETDEWEB)

    Patterson, John [NAC International, Aiken, SC (United States). Foreign Research Reactor Liaison]. E-mail: nacaiken@aol.com

    1998-07-01

    The variations in research reactor fuel form, configuration, irradiation characteristics, and transport cask have required a substantial number of transport cask licensing actions associated with foreign research reactor spent fuel transportation. When compounded by limited time for shipment preparations, due to contract timing or delayed receipt of technical data, the number and timing of certifications has adversely impacted the ability of regulatory agencies to support intended shipping schedules. This issue was brought into focus at a april, 1998 meeting among DOE, the US Nuclear Regulatory Commission, and DOE's spent fuel transportation contractors. (author)

  2. Source storage and transfer cask: Users Guide

    International Nuclear Information System (INIS)

    The storage and shield cask for the dual californium source is designed to shield and transport up to 3.7 mg (2 Ci) of 252Cf. the cask meets Department of Transportation (DOT) license requirements for Type A materials (DOT-7A). The cask is designed to transfer sources to and from the Flourinel and Fuel Storage (FAST) facility delayed-neutron interrogator. Californium sources placed in the cask must be encapsulated in the SR-CF-100 package and attached to Teleflex cables. The cask contains two source locations. Each location contains a gear box that allows a Teleflex cable to be remotely moved by a hand crank into and out of the cask. This transfer procedure permits sources to be easily removed and inserted into the delayed-neutron interrogator and reduces personnel radiation exposure during transfer. The radiation dose rate with the maximum allowable quantity of californium (3.7 mg) in the cask is 30 mR/h at the surface and less than 2 mR/h 1 m from the cask surface. This manual contains information about the cask, californium sources, describes the method to ship the cask, and how to insert and remove sources from the cask. 28 figs

  3. COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

    Energy Technology Data Exchange (ETDEWEB)

    THIELGES, J.R.; CHASTAIN, S.A.

    2007-06-21

    The Disposable Solid Waste Cask (DSWC) is a shielded cask capable of transporting, storing, and disposing of six non-fuel core components or approximately 27 cubic feet of radioactive solid waste. Five existing DSWCs are candidates for use in storing and disposing of non-fuel core components and radioactive solid waste from the Interim Examination and Maintenance Cell, ultimately shipping them to the 200 West Area disposal site for burial. A series of inspections, studies, analyses, and modifications were performed to ensure that these casks can be used to safely ship solid waste. These inspections, studies, analyses, and modifications are summarized and attached in this report. Visual inspection of the casks interiors provided information with respect to condition of the casks inner liners. Because water was allowed to enter the casks for varying lengths of time, condition of the cask liner pipe to bottom plate weld was of concern. Based on the visual inspection and a corrosion study, it was concluded that four of the five casks can be used from a corrosion standpoint. Only DSWC S/N-004 would need additional inspection and analysis to determine its usefulness. The five remaining DSWCs underwent some modification to prepare them for use. The existing cask lifting inserts were found to be corroded and deemed unusable. New lifting anchor bolts were installed to replace the existing anchors. Alternate lift lugs were fabricated for use with the new lifting anchor bolts. The cask tiedown frame was modified to facilitate adjustment of the cask tiedowns. As a result of the above mentioned inspections, studies, analysis, and modifications, four of the five existing casks can be used to store and transport waste from the Interim Examination and Maintenance Cell to the disposal site for burial. The fifth cask, DSWC S/N-004, would require further inspections before it could be used.

  4. COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

    International Nuclear Information System (INIS)

    The Disposable Solid Waste Cask (DSWC) is a shielded cask capable of transporting, storing, and disposing of six non-fuel core components or approximately 27 cubic feet of radioactive solid waste. Five existing DSWCs are candidates for use in storing and disposing of non-fuel core components and radioactive solid waste from the Interim Examination and Maintenance Cell, ultimately shipping them to the 200 West Area disposal site for burial. A series of inspections, studies, analyses, and modifications were performed to ensure that these casks can be used to safely ship solid waste. These inspections, studies, analyses, and modifications are summarized and attached in this report. Visual inspection of the casks interiors provided information with respect to condition of the casks inner liners. Because water was allowed to enter the casks for varying lengths of time, condition of the cask liner pipe to bottom plate weld was of concern. Based on the visual inspection and a corrosion study, it was concluded that four of the five casks can be used from a corrosion standpoint. Only DSWC S/N-004 would need additional inspection and analysis to determine its usefulness. The five remaining DSWCs underwent some modification to prepare them for use. The existing cask lifting inserts were found to be corroded and deemed unusable. New lifting anchor bolts were installed to replace the existing anchors. Alternate lift lugs were fabricated for use with the new lifting anchor bolts. The cask tiedown frame was modified to facilitate adjustment of the cask tiedowns. As a result of the above mentioned inspections, studies, analysis, and modifications, four of the five existing casks can be used to store and transport waste from the Interim Examination and Maintenance Cell to the disposal site for burial. The fifth cask, DSWC S/N-004, would require further inspections before it could be used

  5. Genetics Home Reference: CASK-related intellectual disability

    Science.gov (United States)

    ... Conditions CASK-related intellectual disability CASK-related intellectual disability Enable Javascript to view the expand/collapse boxes. ... Open All Close All Description CASK -related intellectual disability is a disorder of brain development that has ...

  6. Test Plan for Cask Identification Detector

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-29

    This document serves to outline the testing of a Used Fuel Cask Identification Detector (CID) currently being designed under the DOE-NE MPACT Campaign. A bench scale prototype detector will be constructed and tested using surrogate neutron sources. The testing will serve to inform the design of the full detector that is to be used as a way of fingerprinting used fuel storage casks based on the neutron signature produced by the used fuel inside the cask.

  7. Radioactive fuel cask railcar humping study

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, L.T. (comp.)

    1978-01-01

    The response of two radioactive shipping casks due to railroad humping shocks was calculated using a spring-mass model. The two railcars for these casks had different coupling mechanisms and different tiedown arrangements. Humping tests had been performed on one of the railcars (ATMX-600) and the resulting shock spectra was used to adjust the spring-mass model to get matching results. One car (designed for cask shipment) was equipped with Freightmaster E-15 end of car coupler and had about /sup 1///sub 8/ in. free travel of the cask skid relative to the car. The other car (ATMX-600), equipped with Miner RF-333 draft gear, was designed for nuclear weapon shipment and adapted to nuclear waste shipment by fastening the casks to the floor. Both car frames were built by the same manufacturer and are very similar. The response of the casks was put in shock spectra format and a parametric study was performed with various cask weights. Additional studies were done on the effects of fastening the loose cask, and using the Freightmaster end of car coupler on the ATMX car. Half-sine response spectra were overlaid to include the natural frequency of the cask tiedown. The resulting shock amplitude was plotted against the cask weight for each car. The results show a constant acceleration level for all the weights on the car with hydraulic end-of-car coupler which results from constant force at that impact velocity. The cask acceleration can be reduced by fastening it to the car, rather than allowing it to move freely through some small space. This study also shows that the cask response can be optimized on railcars without hydraulic draft gear by adjusting the tiedown stiffness to keep the tiedown frequency different than car frequencies.

  8. Spent fuel shipping cask accident evaluation

    International Nuclear Information System (INIS)

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

  9. Development of cask and transportation system

    International Nuclear Information System (INIS)

    Transportation of spent fuels to the AFR interim storage facility and disposal repository are necessary in Korea. Therefore, an emphasis has been concentrated to develop the design and fabrication technology of commercial casks. A conceptual design of the temperature and deformation measuring systems in the cask, which will be used for mock-up tests has been performed. Preliminary design data of the cask for 7 spent PWR fuels have been obtained in the course of study. (author)

  10. Used Fuel Cask Identification through Neutron Profile

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Laboratory

    2015-11-20

    Currently, most spent fuel is stored near reactors. An interim consolidated fuel storage facility would receive fuel from multiple sites and store it in casks on site for decades. For successful operation of such a facility there is need for a way to restore continuity of knowledge if lost as well as a method that will indicate state of fuel inside the cask. Used nuclear fuel is identifiable by its radiation emission, both gamma and neutron. Neutron emission from fission products, multiplication from remaining fissile material, and the unique distribution of both in each cask produce a unique neutron signature. If two signatures taken at different times do not match, either changes within the fuel content or misidentification of a cask occurred. It was found that identification of cask loadings works well through the profile of emitted neutrons in simulated real casks. Even casks with similar overall neutron emission or average counts around the circumference can be distinguished from each other by analyzing the profile. In conclusion, (1) identification of unaltered casks through neutron signature profile is viable; (2) collecting the profile provides insight to the condition and intactness of the fuel stored inside the cask; and (3) the signature profile is stable over time.

  11. Transportation capabilities of the existing cask fleet

    International Nuclear Information System (INIS)

    This paper describes a number of scenarios estimating the amount of spent nuclear fuel that could be transported to a Monitored Retrievable Storage (MRS) Facility by various combinations of existing cask fleets. To develop the scenarios, the data provided by the Transportation System Data Base (TSDB) were modified to reflect the additional time for cask turnaround resulting from various startup and transportation issues. With these more realistic speed and cask-handling assumptions, the annual transportation capability of a fleet consisting of all of the existing casks is approximately 465 metric tons of uranium (MTU). The most likely fleet of existing casks that would be made available to the DOE consists of two rail, three overweight truck, and six legal weight truck casks. Under the same transportation assumptions, this cask fleet is capable of approximately transporting 270 MTU/year. These ranges of capability is a result of the assumptions pertaining to the number of casks assumed to be available. It should be noted that this assessment assumes additional casks based on existing certifications are not fabricated

  12. Feasibility of dry cask-to-cask and pool-to-cask spent fuel transfer based on single-element transfer cask experience

    International Nuclear Information System (INIS)

    Spent fuel transportation casks and canister-based storage systems are generally loaded underwater in a nuclear plant's spent fuel pool/cask loading pit. Several reasons exist for exploring the feasibility of dry cask-to-cask and pool-to-cask spent fuel transfer. These include: the accommodation of plants which do not have sufficient crane capacity to handle large 90 tonne (100 ton) storage canisters or shipping casks, and construction of an MRS without the need for extensive hot cell facilities. In the case of DOE's ''Multi-Purpose Canister'' (MPC) scenario, use of such a transfer system would allow all plants with adequate transport routes to use large canisters at-reactor, and those without adequate transport routes to use the MRS for loading of large canisters without the need for hot cell facilities. The dry transfer option would not only allow the use of large canisters for all fuel, but would assist DOE in meeting MRS deadlines since licensing and construction of hot-cell facilities significantly affect schedule. This paper reviews the regulatory issues and technical design considerations for a single-element dry transfer system. Also summarized are lessons learned from the TMI-2 fuel transfer system which are directly applicable to the design, testing, startup, and use of a future dry cask-to-cask or pool-to-cask transfer system

  13. Shielding Analysis of the 5320 Shipping Cask

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, A. [Westinghouse Savannah River Company, AIKEN, SC (United States); Nathan, S. [Westinghouse Safety Management Solutions, Aiken, SC (United States)

    1998-05-01

    The purpose of this work is to demonstrate that the 5320 shipping cask meets Federal regulations for maximum radiation dose rates when loaded with the intended plutonium oxide cargo. It should be emphasized that the 5320 is an existing cask, and therefore this work represents confirmatory analysis rather than design analysis.

  14. Design review report FFTF interim storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Scott, P.L.

    1995-01-03

    Final Design Review Report for the FFTF Interim Storage Cask. The Interim Storage Cask (ISC) will be used for long term above ground dry storage of FFTF irradiated fuel in Core Component Containers (CCC)s. The CCC has been designed and will house assemblies that have been sodium washed in the IEM Cell. The Solid Waste Cask (SWC) will transfer a full CCC from the IEM Cell to the RSB Cask Loading Station where the ISC will be located to receive it. Once the loaded ISC has been sealed at the RSB Cask Loading Station, it will be transferred by facility crane to the DSWC Transporter. After the ISC has been transferred to the Interim Storage Area (ISA), which is yet to be designed, a mobile crane will be used to place the ISC in its final storage location.

  15. 78 FR 78165 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-26

    ... RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9... amends the NRC's spent fuel storage regulations by revising the Holtec International HI-STORM 100 Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 9...

  16. Seismic considerations for spent nuclear fuel storage in dry casks

    Institute of Scientific and Technical Information of China (English)

    John L Bignell; Jeffrey A Smith; Christopher A Jones; Susan Y Pickering

    2013-01-01

    To aid the United States Nuclear Regulatory Commission,Sandia National Laboratories (SNL) was contracted to investigate the seismic behavior of typical dry cask storage systems.Parametric evaluations characterized the sensitivity of calculated cask response characteristics to input parameters.The parametric evaluation investigated two generic cask designs (cylindrical and rectangular),three different foundation types (soft soil,hard soil,and rock),and three different casks to pad coefficients of friction (0.2,0.55,0.8) for earthquakes with peak ground accelerations of 0.25g,0.6g,1.0g and 1.25g.A total of 1 165 analyses were completed,with regression analyses being performed on the resulting cask response data to determine relationships relating the mean (16 % and 84 % confidence intervals on the mean) to peak ground acceleration,peak ground velocity,and pseudo-spectral acceleration at 1 Hz and 5 % damping.In general,the cylindrical casks experienced significantly larger responses in comparison to the rectangular cask.The cylindrical cask experienced larger top of cask displacements,larger cask rotations (rocking),and more occurrences of cask toppling (the rectangular cask never toppled over).The cylindrical cask was also susceptible to rolling once rocking had been initiated,a behavior not observed in the rectangular cask.Cask response was not overly sensitive to foundation type,but was significantly dependent on the response spectrum employed.

  17. Selected concrete spent fuel storage cask concepts and the DOE/PSN Cooperative Cask Testing Program

    International Nuclear Information System (INIS)

    To date, water pools, metal casks, horizontal concrete modules, and modular vaults have been used to store the major quantity of commercial light water reactor spent nuclear fuel. Recently, vertical concrete dry storage casks have received consideration for storage of spent nuclear fuel. This paper reviews the evolution of the development of selected vertical concrete dry storage casks and outlines a cooperative cask testing (heat transfer and shielding) program involving the US Department of Energy and Pacific Sierra Nuclear Associates. Others participating in the cooperative program are Pacific Northwest Laboratory; EG ampersand G Idaho, Inc.; Wisconsin Electric Power Company; and the Electric Power Research Institute. 28 refs., 14 figs

  18. Design report for cask transportation equipment

    International Nuclear Information System (INIS)

    In Korea, the spent fuels stored in the spent fuel storage pools in the domestic nuclear power plants significantly affects the continuation of the power plant operation. To solve this problem, KAERI has developed KSC-4 spent fuel shipping cask, which can transport 4 PWR spent fuel assemblies. Besides the development of the cask, KAERI developed transportation equipment which needed to use of KSC-4 cask. These equipment consist of cask handling tools such as lifting yoke, lid handling tool and spent fuel handling tool, etc. and transportation equipment such as trailer. In this report the usages, structures and functions of these tools and equipment were described, and the safety evaluation was carried out for each equipment

  19. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

  20. Dry Cask Storage Characterization Project - Phase 1: CASTOR V/21 Cask Opening and Examination

    International Nuclear Information System (INIS)

    This report documents visual examination and testing conducted in 1999 and early 2000 at the Idaho National Engineering and Environmental Laboratory (INEEL) on a Gesellschaft fuer Nuklear Service (GNS) CASTOR V/21 pressurized water reactor (PWR) spent fuel dry storage cask. The purpose of the examination and testing is to develop a technical basis for renewal of licenses and Certificates of Compliance for dry storage systems for spent nuclear fuel and high-level waste at independent spent fuel storage installation sites. The examination and testing was conducted to assess the condition of the cask internal and external surfaces, cask contents consisting of 21 Westinghouse PWR spent fuel assemblies from Dominion's (formerly named Virginia Power) Surry Power Station and cask concrete storage pad. The assemblies have been continuously stored in the CASTOR cask since 1985. Cask exterior surface and selected fuel assembly temperatures, and cask surface gamma and neutron dose rates were measured. Cask external/internal surfaces, fuel basket components including accessible weldments, fuel assembly exteriors, and primary lid seals were visually examined. Selected fuel rods were removed from one fuel assembly, visually examined, and then shipped to Argonne National Laboratory for nondestructive, destructive, and mechanical examination. Cask interior crud samples and helium cover gas samples were collected and analyzed. The results of the examination and testing indicate the concrete storage pad, CASTOR V/21 cask, and cask contents exhibited sound structural and seal integrity and that long-term storage has not caused detectable degradation of the spent fuel cladding or the release of gaseous fission products between 1985 and 1999

  1. SNS Inner Plug Shipping Cask Analysis

    International Nuclear Information System (INIS)

    Calculations were performed to evaluate the dose rates outside the shipping cask containing the Spallation Neutron Source (SNS) inner plug assembly. The analysis consisted of simulating the proton beam interaction with the SNS target, activation calculations with the determined neutron flux levels and assumed SNS operation schedule, and calculation of the decay gamma-rays propagation through the inner plug and shipping cask. Several materials were considered for the inner plug. The results provide guidance for the finalization of the plug design

  2. Cask Processing Enclosure Specification/Operation - 12231

    International Nuclear Information System (INIS)

    Following an evaluation of throughput rates in the Hot Cell at the Transuranic Waste Processing Center and considering the variability in the waste with respect to actual dose rates a new approach to processing transuranic waste was necessary. Compounding the issue was the remote equipment poor reliability and high down-time. After considering all the factors, the evaluation resulted in the design and construction of a new waste processing area for handling the concrete casks that predominately contain contact handled transuranic (TRU) waste. The area is called the Cask Processing Enclosure and essentially the Cask Processing Enclosure mimics the projects current process techniques used for processing Contact Handled -TRU waste in the existing Box Breakdown Area and Glovebox. The Cask Processing Enclosure approach was developed based on a review of the RH processing throughput rates in the Hot Cell. As the process was reviewed consideration was given to the variability in the waste with respect to actual dose rates and the lack of equipment reliability and high wear in the Hot Cell. Based on that review, a new contact handled processing area for handling the concrete casks is being constructed and startup is expected shortly following WM2012. The Cask Processing Enclosure essentially mimics the projects current process techniques used for processing Contact Handled waste in the existing Box Breakdown Area and Glovebox and the design takes into consideration six years of operational experience. (authors)

  3. 78 FR 63408 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ10 List of Approved Spent Fuel Storage Casks: Transnuclear, Inc.... Nuclear Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising the Transnuclear, Inc. Standardized NUHOMS Cask System listing within the ``List of Approved Spent Fuel...

  4. The impact of using reduced-capacity baskets on cask fleet size and cask fleet mix

    International Nuclear Information System (INIS)

    The Civilian Radioactive Waste Management System transportation system will encounter a wide range of spent fuel characteristics. Since the Initiative I casks are being designed to transport 10-year-old fuel with a burnup of 35,000 MWd/MTU, there is a good likelihood that a number of the cask shipments will need to be derated in order to meet the Nuclear Regulatory Commission radiation guidelines. This report discusses the impact of cask derating by using reduced-capacity baskets. Cask derating, while enhancing the ability to move spent fuel with a wider range of age and burnup characteristics, increases the number of shipments; the amount of equipment (cask bodies, baskets, etc.); and the number of visits to both shipping and receiving sites required to transport a specific amount of spent fuel

  5. Spent Fuel Dry Storage Cask Thermal Test

    International Nuclear Information System (INIS)

    Most nuclear power plants maintain their spent fuel discharged at a reactor in wet storage pools. However, after several years of use, many pools are filled to capacity. Therefore, finding a sufficient capacity for storage is essential because of the continued delays in obtaining a safe, interim storage facility if nuclear power plants are to be allowed to continue to operate. Dry storage cask will be one solution for solving an interim storage problem. The dry storage cask consists of two separate components: an over-pack, and a canister. The structure strength part of the over-pack is made of carbon steel, and the inner cavity of the structure strength part is filled with concrete, which accomplishes the role as a radiation shield. The outer diameter of the dry storage cask is 3,550 mm and the its overall height is 5,885 mm. It weighs approximately 135 tons. The dry storage cask accommodates 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat from the 24 PWR spent fuel assemblies is 25.2 kW This paper discusses the experimental approach used to evaluate the heat transfer characteristics of the dry storage cask

  6. 78 FR 73379 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-06

    ... Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a... of approved spent fuel storage casks in 10 CFR 72.214 (65 FR 25241; May 1, 2000). The environmental... 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9...

  7. Economic evaluation of nuclear waste transportation casks

    International Nuclear Information System (INIS)

    A method is described which allows the systematic economic evaluation of transportation cask designs which meet the requirements of the Test and Evaluation Facility (TEF) program. The heart of the method described is the Waste Management Transportation Model. This model uses a set of computer-based algorithms to assemble specific case information input, combine this input with the data base of transportation information maintained within the model, and calculate the cask types and quantities necessary, the cask utilization factors, and the total costs for each transport line specified. The model is capable of handling a large variety of transportation problems given the specific input related to each type. Three combinations of waste packaging facilities were examined. The first assumes all consolidation and packaging occurs at an existing hot cell. The second assumes all consolidation and packaging is done at the TEF site. The third combination assumes that spent fuels are consolidated at an existing hot cell while waste packaging occurs at the TEF site. Some of the general findings are: (1) defense high-level waste (DHLW) is generally lower in cost than SF as the prime waste form because of the fewer number of shipments required prior to the waste consolidation activity; (2) when DHLW is the prime waste form, it is beneficial to locate the packaging facility (PF) close to the TEF site because the packaged waste form is heavier, more costly to transport; (3) when SF is the prime waste form, it is beneficial to locate the PF close to the waste source to reduce the length of the transport links containing unconsolidated spent fuel assemblies; and (4) truck casks, and legal weight truck casks in particular, are generally superior to the rail casks on an economic basis

  8. A numerical study of transportation casks subjected to puncture loads

    International Nuclear Information System (INIS)

    A nonlinear dynamic finite element analysis has been performed to study the structural response of casks subjected to puncture load. Particular attention is placed on the Multipurpose Canister (MPC) and General Atomic (GA) casks that are currently under development. The structural response of the casks subjected to both regulatory hypothetical accidents and accidents beyond regulatory requirements were evaluated. A performance map was presented for casks subjected to regulatory formula puncture tests, and the structural contribution of the various layers backing the steel cask shell has been studied

  9. Safety evaluation for packaging (onsite) SERF cask

    International Nuclear Information System (INIS)

    This safety evaluation for packaging (SEP) documents the ability of the Special Environmental Radiometallurgy Facility (SERF) Cask to meet the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B quantities (up to highway route controlled quantities) of radioactive material within the 300 Area of the Hanford Site. This document shall be used to ensure that loading, tie down, transport, and unloading of the SERF Cask are performed in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required

  10. Safety evaluation for packaging (onsite) SERF cask

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.S.

    1997-10-24

    This safety evaluation for packaging (SEP) documents the ability of the Special Environmental Radiometallurgy Facility (SERF) Cask to meet the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B quantities (up to highway route controlled quantities) of radioactive material within the 300 Area of the Hanford Site. This document shall be used to ensure that loading, tie down, transport, and unloading of the SERF Cask are performed in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  11. The Feasibility of Cask "Fingerprinting" as a Spent-Fuel, Dry-Storage Cask Safeguards Technique

    Energy Technology Data Exchange (ETDEWEB)

    Ziock, K P; Vanier, P; Forman, L; Caffrey, G; Wharton, J; Lebrun, A

    2005-07-27

    This report documents a week-long measurement campaign conducted on six, dry-storage, spent-nuclear-fuel storage casks at the Idaho National Laboratory. A gamma-ray imager, a thermal-neutron imager and a germanium spectrometer were used to collect data on the casks. The campaign was conducted to examine the feasibility of using the cask radiation signatures as unique identifiers for individual casks as part of a safeguards regime. The results clearly show different morphologies for the various cask types although the signatures are deemed insufficient to uniquely identify individual casks of the same type. Based on results with the germanium spectrometer and differences between thermal neutron images and neutron-dose meters, this result is thought to be due to the limitations of the extant imagers used, rather than of the basic concept. Results indicate that measurements with improved imagers could contain significantly more information. Follow-on measurements with new imagers either currently available as laboratory prototypes or under development are recommended.

  12. CASKS (Computer Analysis of Storage casKS): A microcomputer based analysis system for storage cask design review. User's manual to Version 1b (including program reference)

    International Nuclear Information System (INIS)

    CASKS (Computer Analysis of Storage casKS) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for evaluating safety analysis reports on spent-fuel storage casks. The bulk of the complete program and this user's manual are based upon the SCANS (Shipping Cask ANalysis System) program previously developed at LLNL. A number of enhancements and improvements were added to the original SCANS program to meet requirements unique to storage casks. CASKS is an easy-to-use system that calculates global response of storage casks to impact loads, pressure loads and thermal conditions. This provides reviewers with a tool for an independent check on analyses submitted by licensees. CASKS is based on microcomputers compatible with the IBM-PC family of computers. The system is composed of a series of menus, input programs, cask analysis programs, and output display programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests

  13. Simplified computer codes for cask impact analysis

    International Nuclear Information System (INIS)

    In regard to the evaluation of the acceleration and deformation of casks, the simplified computer codes make analyses economical and decrease input and calculation time. The results obtained by the simplified computer codes have enough adequacy for their practical use. (J.P.N.)

  14. Surface storage cask test summarization report

    International Nuclear Information System (INIS)

    From December 1978 to September 1982, as part of DOE's Spent Fuel Handling and Packaging Program and Commercial Waste and Spent Fuel Packaging Program, a pressurized water reactor (PWR) spent nuclear fuel assembly with an initial decay heat level of approximately 1.0 kilowatt (kW) was emplaced in a concrete cask at the Engine Maintenance, Assembly and Disassembly (E-MAD) facility in Area 25 of the Nevada Test Site. Temperatures were monitored during the emplacement period to determine the thermal response of the cask, the canister, and the fuel assembly. During and following the test, the atmosphere of the canister containing the fuel assembly was sampled to determine if fission product gases had been released by the fuel assembly. This 45-month Surface Storage Cask (SSC) test was the first demonstration of interim storage of a PWR spent fuel assembly in a dry storage cask. The receipt, handling, packaging, emplacement and retrieval operations have been demonstrated as directly applicable to similar operations in federal interim storage and repository related activities. 7 references, 35 figures, 7 tables

  15. Evaluation of improvement potential for spent fuel cask handling

    International Nuclear Information System (INIS)

    This report describes the quantitative analysis of opportunities to improve the loading/unloading operations for spent fuel shipping casks. The improvement potential is defined as a reduction in the time for completion or worker exposure for the complete handling operations. Two casks have been chosen as representative of presently available shipping casks. These are the NAC-1/NFS-4 legal weight truck cask and the IF-300 rail cask. The handling operations for each of these casks are broken down into a series of sequential steps. The time for completion and worker exposure is described by a probability density function for each step. These step descriptions are then combined to form a base case description of the total loading/unloading operation. Potential improvement opportunities are evaluated by modifying the appropriate probability density function descriptors then recombining the steps to form a probabilistic description of the modified operation

  16. Safety analysis report for EPMA irradiated specimen cask

    Energy Technology Data Exchange (ETDEWEB)

    Ku, J. H.; Lee, J. C.; Seo, K. S.; Bang, K. S.; Park, S. W.; Min, D. K

    2000-11-01

    For the effective examination of spent fuels and radioactive materials by using EPMA in IMEF besides using SEM in PIEF, a special purpose EPMA cask was developed. It will be used to transport a specimen from the hot-cell in PIEF to the shielded glove box in IMEF. This cask should be easy to handle and transport by hand carry. It also has to be safe to maintain the shielding safety as well as the thermal and structural integrities under prescribed load conditions by the regulatory requirements. This cask was designed compactly to be docked perfectly maintaining shielding integrity without the modification of the interfaces of hot-cell and shielded glove box. Accordingly, the main features of cask were analyzed with functional capabilities, and the integrities of cask under required load conditions were evaluated. It was verified that the EPMA cask is suitable to use at handy transport of irradiated specimen between the PIEF and IMEF facilities in KAERI.

  17. Interfacing the existing cask fleet with the MRS

    International Nuclear Information System (INIS)

    This paper reports that the Department of Energy (DOE) is considering the possibility of using the existing fleet of casks to achieve spent fuel receipt at the Monitored Retrievable Storage (MRS) facility. The existing cask fleet includes the NLI-1/2, the NAC-LWT, the TN-8 (and TN-8L), the TN-9, and the IF-300 casks. Other casks may be available, but their status is not certain. Use of the existing cask fleet at the MRS places additional design requirements on the system, and specifically affects the cask-to-MRS interface. The decision to use the existing cask fleet also places additional demands on training needs and operator certification, and the configuration management system. Some existing cask designs may not be able to mate with a bottom opening hot cell MRS. Use of the existing cask fleet also greatly increases the number of shipments that must be received, to the point that a facility larger than originally envisioned may be required

  18. Studies and research concerning BNFP: cask handling equipment standardization

    International Nuclear Information System (INIS)

    This report covers the activities of one of the sub-tasks within the Spent LWR Fuel Transportation Receiving, Handling, and Storage program. The sub-task is identified as Cask Handling Equipment Standardization. The objective of the sub-task specifies: investigate and identify opportunities for standardization of cask interface equipment. This study will examine the potential benefits of standardized yokes, decontamination barriers and special tools, and, to the extent feasible, standardized methods and software for handling the variety of casks presently available in the US fleet. The result of the investigations is a compilation of reports that are related by their common goal of reducing cask turnaround time

  19. Rail tiedown tests with heavy casks for radioactive shipments

    International Nuclear Information System (INIS)

    A rail tiedown test program was conducted at the Savannah River Plant in July and August 1978. For each test, a 40- or 70-ton cask was secured on a railcar. The railcar was pushed to speeds up to 11 mph and allowed to couple to parked railcars simulating ordinary railyard operations. The test car carrying the cask was heavily instrumented to measure the accelerations and forces generated at strategically selected places. Eighteen test runs were made with different combinations of railcars, couplers, casks, speeds, and tiedown configurations. The major objectives of the test program were to (1) provide test data as a basis to develop a tiedown standard for rail cask shipments of radioactive materials and (2) collect dynamic data to support analytical models of the railcar cask tiedown system. The optimum tiedown configuration demonstrated for heavy casks was a combination of welded, fixed stops to secure the cask longitudinally and flexible cables to restrain vertical and lateral cask movement. Cables alone were inadequate to secure a heavy cask to a standard railcar, and bolting was found disadvantageous in several respects. The use of cushioning coupler mechanisms dramatically reduced the tiedown requirements for the rail coupling operation. The test program and general conclusions are discussed

  20. Inspection of Used Fuel Dry Storage Casks

    Energy Technology Data Exchange (ETDEWEB)

    Dennis C. Kunerth; Tim McJunkin; Mark McKay; Sasan Bakhtiari

    2012-09-01

    ABSTRACT The U.S. Nuclear Regulatory Commission (NRC) regulates the storage of used nuclear fuel, which is now and will be increasingly placed in dry storage systems. Since a final disposition pathway is not defined, the fuel is expected to be maintained in dry storage well beyond the time frame originally intended. Due to knowledge gaps regarding the viability of current dry storage systems for long term use, efforts are underway to acquire the technical knowledge and tools required to understand the issues and verify the integrity of the dry storage system components. This report summarizes the initial efforts performed by researchers at Idaho National Laboratory and Argonne National Laboratory to identify and evaluate approaches to in-situ inspection dry storage casks. This task is complicated by the design of the current storage systems that severely restrict access to the casks.

  1. Drop test of transportable storage cask

    International Nuclear Information System (INIS)

    It is being planned to transport the transportable storage casks again after their storage period of several decades, so metal gaskets are used as seal material in their lids in place of rubber o-rings which deteriorate during the storage period. Since the slightest dislocation of the lids causes seal performance deterioration in the metal gaskets, it is necessary to establish a simulation technology which accurately estimates the dislocation in order to design a rigid lid structure to protect against the impact loads under 9 m drop condition. A 1:3 scale model of the transportable storage cask developed by Hitz for BWR spent fuel rods were manufactured and 9 m drop tests were performed. Measured dislocations of the lids were confirmed within the allowable limit and they were found to be accurately simulated. (author)

  2. MCO loading and cask loadout technical manual

    Energy Technology Data Exchange (ETDEWEB)

    PRAGA, A.N.

    1998-10-01

    A compilation of the technical basis for loading a multi-canister overpack (MCO) with spent nuclear fuel and then placing the MCO into a cask for shipment to the Cold Vacuum Drying Facility. The technical basis includes a description of the process, process technology that forms the basis for loading alternatives, process control considerations, safety considerations, equipment description, and a brief facility structure description.

  3. Concrete Spent Fuel Cask Criticality Calculation

    International Nuclear Information System (INIS)

    A preliminary analysis of the concrete cask for the intermediate dry storage of the spent fuel of NPP Krsko should include an estimation of the effective multiplication factor. Assuming 16x16 fuel elements, 4.3% initial enrichment, 45 GWd/tU burnup and 10 years cooling time, a concrete spent fuel capacity of 10 spent fuel assemblies is proposed. Fuel assemblies are placed inside inner cavity in a 'basket' - a boron (1%) doped steel structure. Heavy concrete (25% Fe), 45 cm thick, is enclosed in a carbon steel shell. There is also a stainless steel (SS304) lining of the storage cavity. Isotope inventory of the spent fuel after a 10 years cooling time is calculated using ORIGEN-S functional module of the SCALE-4.2 code package. The effective multiplication factor keff of dry (helium filled) and wet (water filled) cask for fresh and used fuel is calculated using CSAS4 Monte Carlo method based control module of the same SCALE-4.2 code package. The obtained results of keff of the dry cask for fresh and spent fuel are well below the required 0.95 value, but those for the water filled cask are above this value. Therefore, several additional calculations of the keff varying the thickness of a boral basket structure which had replaced the stainless steel one were done. It turned out that at least a 1.5 cm thick boral wall was needed to meet the required 0.95 value for keff. (author)

  4. ANALISIS LITERASI EKONOMI

    Directory of Open Access Journals (Sweden)

    Peter Garlans Sina

    2012-10-01

    Full Text Available Abstract: Analysis of Economic Literacy. The aim of this research is as an effort to increase economic literacy for individuals and households who want to achieve prosperity. Therefore, the obligation of improving the economic literacy needs to be done in a well-planned manner and come from the strong intention to learn in order to improve the economic literacy. It could have an effect on the assets accumulation, a better debt management, as a protection, as well as to increase savings and managing spending intelligently.   Keyword: economic literacy, asset, debt, protection, saving, spending   Abstrak: Analisis Literasi Ekonomi. Tulisan ini bertujuan sebagai upaya meningkatkan literasi ekonomi bagi individu maupun rumah tangga yang menginginkan mencapai kesejahteraan. Oleh karena itu, kewajiban meningkatkan literasi ekonomi perlu dilakukan secara terencana dan diawali dari niat untuk belajar meningkatkan literasi ekonomi karena dapat berefek pada akumulasi aset, pengelolaan utang yang tepat, proteksi, meningkatkan tabungan dan cerdas mengelola pengeluaran.   Kata kunci: literasi ekonomi, aset, utang, proteksi, menabung, pengeluaran

  5. Safety analysis report for packaging: the ORNL loop transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Nelms, H.A.; Crowley, W.K.; Just, R.A.

    1977-11-01

    An evaluation of the ORNL loop transport cask demonstrating its compliance with the regulations governing the transportation of radioactive and fissile materials is presented. A previous review of the cask is updated to demonstrate compliance with current regulations, to present current procedures, and to reflect the more recent technology.

  6. 78 FR 73456 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-06

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI.... SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its spent fuel storage... Approved Spent Fuel Storage Casks'' to include Amendment No. 9 to Certificate of Compliance (CoC) No....

  7. 78 FR 78285 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-26

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI... public comment period. The document proposed to amend the NRC's spent fuel storage regulations by revising the Holtec International HI-STORM 100 Cask System listing within the ``List of Approved Spent...

  8. US cask requirements and industry capability survey

    International Nuclear Information System (INIS)

    The objectives of this paper are to provide an estimate of spent fuel shipping cask requirements for reactor to away-from-reactor (AFR) storage facility shipments from the present time until late in this century and to determine and document the willingness and capability of private industry to provide required future transportation services. In order to meet this objective, the Transportation Technology Center at Sandia National Laboratories sponsored Teledyne Energy Systems to conduct a survey of US industry. Results of tasks completed to carry out the objectives are reviewed

  9. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    International Nuclear Information System (INIS)

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented

  10. DOE procurement activities for spent fuel shipping casks

    International Nuclear Information System (INIS)

    This paper discusses the DOE cask development program established to satisfy the requirements of the NWPA. The program is designed to provide safe efficient casks on a timely schedule. The casks will be certified by the NRC in compliance with the 1987 amendment to NWPA. Private industry will be used to the maximum extent. DOE will encourage use of present cask technology, but will not hesitate to advance the state-of-the-art to improve efficiency in transport operations, provided that safety is not compromised. DOE will support the contractor's efforts to advance the state-of-the-art by maintaining a technical development effort that will respond to the common needs of all the contractors. DOE and the cask contractors will develop comprehensive and well integrated programs of test and analysis for cask certification. Finally, the DOE will monitor the cask development program within a system that fosters early identification of improvement opportunities as well as potential problems, and is sufficiently flexible to respond quickly yet rationally to assure a fully successful program

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

    International Nuclear Information System (INIS)

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

  12. Differences of Technical Requirements Between Transportation and Storage Metal Casks

    International Nuclear Information System (INIS)

    The worldwide demand of storage facilities for spent fuels discharged from nuclear power stations is increasing to maintain sustainable operation of the nuclear power stations. The spent fuels are stored at first in the fuel pools (wet storage). When the spent fuels exceed the pool storage capacity, the fuels are transferred to the other storage facility located at reactor or away from reactor, which often adopts a dry storage technology. To use metal casks is one of the options for the dry storage facilities, and some storage facilities have already utilized large metal casks, whose original design concept were developed to transport the spent fuels from nuclear power stations to a reprocessing plant by trains, trucks or by sea-going vessels. It is widely understood that the technology of transportation casks developed up to now is able to apply to the storage casks without any significant design changes. Technical requirements on the design are discussed between the storage cask and the transportation cask to confirm of the understanding based on the assumption that the large metal cask is used for transportation and storage respectively. (author)

  13. FACSIM/MRS-1: Cask receiving and consolidation performance assessment

    International Nuclear Information System (INIS)

    A simulation analysis was completed to assess the performance of the shipping cask receiving and spent-fuel handling, consolidation and canistering operations of the Monitored Retrievable Storage (MRS) facility. One purpose of this evaluation was to estimate the limits of MRS operational capabilities and factors leading to those limitations. The model used to obtain the performance assessment, FACSIM/MRS-1, is one of two components of the FACSIM model developed by PNL's simulation effort for the nuclear waste-handling facility. FACSIM/MRS-1 provides the user with information about lag-storage requirements, machine use, cask queues, welder queues, and cask process and cask turnaround times. The model can help determine the effect that the following activities have on operating efficiency: (1) receiving multiple cask shipments, when rail-cask or truck-cask shipments arrive at the facility in groups of two or more, and (2) operating the facility five days per week, three shifts per day or seven days per week, three shifts per day for any conditions. In addition, sensitivity to equipment failure frequency and the time needed for equipment repair can be studied. Information on the above operating characteristics may be obtained for any spent-fuel rate, any split of shipments between truck and rail transport, or any split of boiling water reactor/pressurized water reactor fuel

  14. Robotic radiation survey and analysis system for radiation waste casks

    International Nuclear Information System (INIS)

    Sandia National Laboratories (SNL) and the Hanford Engineering Development Laboratories have been involved in the development of remote systems technology concepts for handling defense high-level waste (DHLW) shipping casks at the waste repository. This effort was demonstrated the feasibility of using this technology for handling DHLW casks. These investigations have also shown that cask design can have a major effect on the feasibility of remote cask handling. Consequently, SNL has initiated a program to determine cask features necessary for robotic remote handling at the waste repository. The initial cask handling task selected for detailed investigation was the robotic radiation survey and analysis (RRSAS) task. In addition to determining the design features required for robotic cask handling, the RRSAS project contributes to the definition of techniques for random selection of swipe locations, the definition of robotic swipe parameters, force control techniques for robotic swipes, machine vision techniques for the location of objects in 3-D, repository robotic systems requirements, and repository data management system needs

  15. Signatures of Extended Storage of Used Nuclear Fuel in Casks

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-28

    As the amount of used nuclear fuel continues to grow, more and more used nuclear fuel will be transferred to storage casks. A consolidated storage facility is currently in the planning stages for storing these casks, where at least 10,000 MTHM of fuel will be stored. This site will have potentially thousands of casks once it is operational. A facility this large presents new safeguards and nuclear material accounting concerns. A new signature based on the distribution of neutron sources and multiplication within casks was part of the Department of Energy Office of Nuclear Energy’s Material Protection, Account and Control Technologies (MPACT) campaign. Under this project we looked at fingerprinting each casks neutron signature. Each cask has a unique set of fuel, with a unique spread of initial enrichment, burnup, cooling time, and power history. The unique set of fuel creates a unique signature of neutron intensity based on the arrangement of the assemblies. The unique arrangement of neutron sources and multiplication produces a reliable and unique identification of the cask that has been shown to be relatively constant over long time periods. The work presented here could be used to restore from a loss of continuity of knowledge at the storage site. This presentation will show the steps used to simulate and form this signature from the start of the effort through its conclusion in September 2016.

  16. Nuclear cask testing films misleading and misused

    Energy Technology Data Exchange (ETDEWEB)

    Audin, L. [Audin (Lindsay), Ossining, NY (United States)

    1991-10-01

    In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as ``proof`` to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.

  17. Nuclear cask testing films misleading and misused

    Energy Technology Data Exchange (ETDEWEB)

    Audin, L. (Audin (Lindsay), Ossining, NY (United States))

    1991-10-01

    In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as proof'' to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.

  18. Nuclear cask testing films misleading and misused

    International Nuclear Information System (INIS)

    In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as ''proof'' to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests

  19. Viability of Existing INL Facilities for Dry Storage Cask Handling

    Energy Technology Data Exchange (ETDEWEB)

    Bohachek, Randy; Wallace, Bruce; Winston, Phil; Marschman, Steve

    2013-04-30

    This report evaluates existing capabilities at the INL to determine if a practical and cost effective method could be developed for opening and handling full-sized dry storage casks. The Idaho Nuclear Technology and Engineering Center (INTEC) CPP-603, Irradiated Spent Fuel Storage Facility, provides the infrastructure to support handling and examining casks and their contents. Based on a reasonable set of assumptions, it is possible to receive, open, inspect, remove samples, close, and reseal large bolted-lid dry storage casks at the INL. The capability can also be used to open and inspect casks that were last examined at the TAN Hot Shop over ten years ago. The Castor V/21 and REA-2023 casks can provide additional confirmatory information regarding the extended performance of low-burnup (<45 GWD/MTU) used nuclear fuel. Once a dry storage cask is opened inside CPP-603, used fuel retrieved from the cask can be packaged in a shipping cask, and sent to a laboratory for testing. Testing at the INL’s Materials and Fuels Complex (MFC) can occur starting with shipment of samples from CPP-603 over an on-site road, avoiding the need to use public highways. This reduces cost and reduces the risk to the public. The full suite of characterization methods needed to establish the condition of the fuel exists and MFC. Many other testing capabilities also exist at MFC, but when those capabilities are not adequate, samples can be prepared and shipped to other laboratories for testing. This report discusses how the casks would be handled, what work needs to be done to ready the facilities/capabilities, and what the work will cost.

  20. Viability of Existing INL Facilities for Dry Storage Cask Handling

    Energy Technology Data Exchange (ETDEWEB)

    Randy Bohachek; Charles Park; Bruce Wallace; Phil Winston; Steve Marschman

    2013-04-01

    This report evaluates existing capabilities at the INL to determine if a practical and cost effective method could be developed for opening and handling full-sized dry storage casks. The Idaho Nuclear Technology and Engineering Center (INTEC) CPP-603, Irradiated Spent Fuel Storage Facility, provides the infrastructure to support handling and examining casks and their contents. Based on a reasonable set of assumptions, it is possible to receive, open, inspect, remove samples, close, and reseal large bolted-lid dry storage casks at the INL. The capability can also be used to open and inspect casks that were last examined at the TAN Hot Shop over ten years ago. The Castor V/21 and REA-2023 casks can provide additional confirmatory information regarding the extended performance of low-burnup (<45 GWD/MTU) used nuclear fuel. Once a dry storage cask is opened inside CPP-603, used fuel retrieved from the cask can be packaged in a shipping cask, and sent to a laboratory for testing. Testing at the INL’s Materials and Fuels Complex (MFC) can occur starting with shipment of samples from CPP-603 over an on-site road, avoiding the need to use public highways. This reduces cost and reduces the risk to the public. The full suite of characterization methods needed to establish the condition of the fuel exists and MFC. Many other testing capabilities also exist at MFC, but when those capabilities are not adequate, samples can be prepared and shipped to other laboratories for testing. This report discusses how the casks would be handled, what work needs to be done to ready the facilities/capabilities, and what the work will cost.

  1. What drives Greek consumer preferences for cask wine?

    DEFF Research Database (Denmark)

    Chrysochou, Polymeros; Corsi, A. M.; Krystallis Krontalis, Athanasios

    2012-01-01

    a sustainable eco-friendly positioning. Originality/value – This study contributes to the understanding of what drives consumers’ preferences for cask wine, something that few studies have done until now. Moreover, this is the first study to use the BWS method for this type of product.......Purpose – Cask wine (bag-in-box, soft pack) has not received considerable attention in wine marketing research, but interest among winemakers and consumers has been increasing steadily. However, little is known about what drives consumer preferences for cask wine and, furthermore, what the profile...

  2. Development of dual-purpose metal cask for interim storage of spent nuclear fuel (1). Outline of cask structure

    International Nuclear Information System (INIS)

    Spent fuels discharged from nuclear power plants in Japan are planed to be reprocessed at the nuclear fuel recycle plant under construction at Rokkasho-mura. Since the amount of the spent fuels exceeds that of recycled fuel, the spent fuels have to be properly stored and maintained as recycle fuel resource until the beginning of the reprocessing. For that sake, interim storage installations are being constructed outside the nuclear power plants by 2010. The storage dry casks have been practically used as the interim storage in the nuclear power plants. From this reason, the storage system using the storage dry casks is promising as the interim storage installations away form the reactors, which are under discussion. In the interim storage facilities, the storage using the dry cask of the storage metal cask with business showings, having the function of transportation is now under discussion. By employing transportation and storage dual-purpose cask, the repack equipments can be exhausted, and the reliability of the interim storage installations can be increased. Hitachi, Ltd. has been developing the high reliable and economical transportation and storage dry metal cask. In this report, the outline of our developing transportation and storage dry cask is described. (author)

  3. ANALISIS PENILAIAN PRESTASI KERJA PEGAWAI

    Directory of Open Access Journals (Sweden)

    Puspita Rokhmawati

    2013-03-01

    Full Text Available Tujuan analisis ini untuk mengetahui sistem, evaluasi dan perubahan penilaian prestasi kerja pegawai UPT Balai Konservasi Tumbuhan Kebun Raya Cibodas -LIPI. Pengambilan sampel dalam penelitian ini dilakukan secara acak dengan metode yang digunakan terdiri dari beberapa bagian, yaitu studi pustaka, studi lapangan dan metode analisis. Metode yang digunakan adalah Metode Analisis SWOT. Hasil analisis menunjukkan banyak kelemahan yang muncul daripada kekuatan yang dimiliki. Peluang dan ancaman yang ada pun tidak memotivasi dalam mengimplementasi ke sistem penilaian prestasi kerja. Oleh karena itu perlu upaya mengevaluasi dan mengubah sistem penilaian prestasi kerja pegawai. Pemberian umpan balik akan mendorong kearah kemajuan dan kemungkinan meningkatkan kualitas kerja pegawai. Maka penilaian harus dibuat seobjektif dan seteliti mungkin berdasarkan data yang tersedia.The purpose of this study is  to determine the system, the  evaluation and the assessment of employees’ job performance in UPT Botanical Gardens Plant Conservation Center Cibodas - LIPI. UPT Botanical Gardens Plant Conservation Center Cibodas - LIPI has not carry out  an optimal job performance assessment. There are indications that the condition relate to the presence of subjectivity in the assessment so depth analysis is needed to find out the solution. Sampling technique uses random sampling. SWOT analysis is used to analyze the data along with study of literature and field study. The result shows that more weaknesses reveale in job performance assessment system than strengths.Opportunities and treats are ignored in implementing the job performance assessment system. Evaluation and modification are needed to improve employees’ job quality.

  4. Country report France [Operation and maintenance of spent fuel storage and transportation casks/containers

    International Nuclear Information System (INIS)

    Transportation from Electricite de France and other foreign utilities to COGEMA La Hague reprocessing plant is performed with one family of casks in the 100 ton class. The experience gained in transport cask design and operation has resulted in design of transport/storage and storage only systems. There are 6 cask types for transportation only and 10 cask types for dual purpose storage and transportation. French authorities approve each cask design. Cask vendors provide training and assistance to users as well as a transportation file containing all actions and recording inspections of the cask. Maintenance frequencies are determined according to design an experience and maintenance specifications prepared. The extent of maintenance is at three levels: inspections on arrival and departure, every 3 years or 15 transports and every 6 years or 60 transports. According to French experience the cask maintenance costs over lifetime are the same as the cost of the cask itself. (author)

  5. Development of tipping-over analysis of cask subjected to earthquake strong motion

    International Nuclear Information System (INIS)

    Since a cask is vertically oriented during loading in cask-storage, it is necessary to investigate the integrity of the cask against tipping-over during strong earthquakes. The rocking and sliding behavior of the cask during strong earthquakes can be analyzed as a dynamic vibration problem for a rigid cylinder. In this paper, in order to clarify the tipping-over characteristics of a cask during strong earthquakes, the authors applied the Distinct Element Method (DEM) to the seismic response analysis of the cask. DEM was introduced by Cundall P.A. in 1971. It is based on the use of an explicit numerical scheme. The cask was considered to be a rigid polygonal element, which satisfied the equation of motion and the law of action and reaction. They examined the applicability of this code by comparison with experimental results obtained from shaking table tests using scale model casks considering the dimension of a 100 ton class full-scale cask

  6. Thermal test and analysis of a spent fuel storage cask

    International Nuclear Information System (INIS)

    A thermal test simulated with full-scale cask model for the normal storage was performed to verify the storage skill of the spent fuels of the cask. The maximum temperature at each point in the test was lower than the allowable temperature. The integrity of the cask was maintained. It was observed that the safety of containment system was also kept according to the check of the seal before and after the thermal test. Therefore it was shown that using the present skill, it is possible to store spent fuels in the dry-type cask safely. Moreover, because of the good agreement between analysis and experimental results, it was shown that the analysis model was successfully established to estimate the temperature distribution of the fuel cladding and the seal portion. (J.P.N.)

  7. Feasibility study for a transportation operations system cask maintenance facility

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the development of a waste management program for the disposition of spent nuclear fuel (SNF) and high-level waste (HLW). The program will include a transportation system for moving the nuclear waste from the sources to a geologic repository for permanent disposal. Specially designed casks will be used to safely transport the waste. The cask systems must be operated within limits imposed by DOE, the Nuclear Regulatory Commission (NRC), and the Department of Transportation (DOT). A dedicated facility for inspecting, testing, and maintaining the cask systems was recommended by the General Accounting Office (in 1979) as the best means of assuring their operational effectiveness and safety, as well as regulatory compliance. In November of 1987, OCRWM requested a feasibility study be made of a Cask Maintenance Facility (CMF) that would perform the required functions. 46 refs., 16 figs., 13 tabs

  8. Feasibility study for a transportation operations system cask maintenance facility

    Energy Technology Data Exchange (ETDEWEB)

    Rennich, M.J.; Medley, L.G.; Attaway, C.R.

    1991-01-01

    The US Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the development of a waste management program for the disposition of spent nuclear fuel (SNF) and high-level waste (HLW). The program will include a transportation system for moving the nuclear waste from the sources to a geologic repository for permanent disposal. Specially designed casks will be used to safely transport the waste. The cask systems must be operated within limits imposed by DOE, the Nuclear Regulatory Commission (NRC), and the Department of Transportation (DOT). A dedicated facility for inspecting, testing, and maintaining the cask systems was recommended by the General Accounting Office (in 1979) as the best means of assuring their operational effectiveness and safety, as well as regulatory compliance. In November of 1987, OCRWM requested a feasibility study be made of a Cask Maintenance Facility (CMF) that would perform the required functions. 46 refs., 16 figs., 13 tabs.

  9. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  10. Operating Experience in Spent Fuel Storage Casks

    International Nuclear Information System (INIS)

    A safe storage of spent fuels has been considered as one of the inevitable tasks for TEPCO for the last few decades. In order to increase flexibility for the fuel storage measures, TEPCO has been storing spent fuels in an on-site dry storage facility at Fukushima-Daiichi Nuclear Power Station. Since 1995, more than 400 fuel assemblies have been safely store. Integrity of storage casks and fuels were carefully checked by periodical inspections, which were conducted in 2000 and 2005. The next investigation will be held within a few years in order to verify the safety conditions even after a 15-year storage. These series of inspections will give plenty of useful data for the design and operation of the Mutsu facility, which will be the first off-site interim spent fuel storage facility away from any reactor site in Japan. (author)

  11. Impact velocity vs. target hardness relationships for equivalent response of cask structures

    International Nuclear Information System (INIS)

    In this paper, impact velocity vs. target hardness relationships for cask structures are reviewed. The relationships are based on equivalent cask responses in terms of equal deceleration or similar cask damages. By examining several past cask or container tests as well as some analytical results, some conclusions can be drawn about the relationship between target hardness and equivalent impact velocities. This relationship clearly shows that the cask response to impact is cask-dependent and that the rigid sphere impact model results in an unconservative estimate of equivalent velocity

  12. Experience with certifying borated stainless steel as a shipping cask basket material

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, D.G. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Nickell, R.E. (Applied Science and Technology, Inc., Poway, CA (USA))

    1990-01-01

    The original cask designs for a cask demonstration project featured fuel baskets constructed of borated stainless steel (bss) as a structural material. The project is intended to demonstrate casks that can be used for both shipping and storing spent nuclear fuel assemblies. The baskets were intended to maintain the fuel assemblies in a subcritical array for both normal and accident conditions. The Nuclear Regulatory Commission, however, judged bss to be unacceptable as a structural material. The cask designs were subsequently modified. The knowledge gained during this cask demonstration project may be applicable to development of bss as a basket material in future cask design. 6 refs., 2 figs., 2 tabs.

  13. Experience with certifying borated stainless steel as a shipping cask basket material

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, D.G. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Nickell, R.E. [Applied Science and Technology, Poway, CA (United States)

    1990-10-01

    This paper discusses the original cask designs for a cask demonstration project that has featured fuel baskets constructed of borated stainless steel (bss) as a structural material. The project is intended to demonstrate casks that can be used for both shipping and storing spent nuclear fuel assemblies. The baskets were intended to maintain the fuel assemblies in a subcritical array for both normal and accident conditions. The Nuclear Regulatory Commission, judged bss to be unacceptable as a structural material. The cask designs were subsequently modified. The knowledge gained during this cask demonstration project may be applicable to development of bss as a basket material in future cask design.

  14. Geometric feasibility of flexible cask transportation system for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lima, P.; Ribeiro, M.I.; Aparicio, P. [Instituto Superior Tecnico-Instituto de Sistemas e Robotica, Lisboa (Portugal)

    1998-07-01

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  15. Geometric feasibility of flexible cask transportation system for ITER

    International Nuclear Information System (INIS)

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  16. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    International Nuclear Information System (INIS)

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior

  17. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    Energy Technology Data Exchange (ETDEWEB)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  18. An economic evaluation of a storage system for casks with burnup credit

    International Nuclear Information System (INIS)

    It is generally recognized that casks designed with burnup credit are more economical than those without burnup credit. To estimate how much more economical they are, we made conceptual designs of transport/storage casks with and without burnup credit for PWR and BWR fuels of various uranium enrichment. The casks were designed to contain the maximum number of fuel assemblies under the necessary weight and dimensional limitations as well as the criticality and shielding criteria. The results showed that approximately 8 % to 44 % more fuel assemblies could be contained in casks with burnup credit. We then evaluated the economy of cask storage system incorporating the cask designs obtained above both with and without burnup credit. The results showed that the cost of storing casks with burnup credit is approximately 7 % to 30 % less expensive than storing casks without burnup credit. (J.P.N.)

  19. Analisi SWOT Terhadap Manajemen Rumah Sakit

    OpenAIRE

    Rahmati MR

    2008-01-01

    SWOT adalah singkatan dari strengths dan weaknesses serta opportunities dan threats. Analisis SWOT merupakan suatu kajian terhadap kekuatan dan kelemahan internal sebuah organisasi serta kesempatan dan ancaman lingkungan eksternalnya. Teknik analisis SWOT yang digunakan adalah mathematics of SWOT yaitu dengan menggunakan rumus (S-W) + (0—T), jika hasil yang diperoleh plus maka strateginya adalah menguatkan S dan O, jika equal maka strateginya adalah mengurangi W dan T9 dan jika minus maka ...

  20. Seismic tipping analysis of a spent nuclear fuel shipping cask sitting on a crush pad

    International Nuclear Information System (INIS)

    A crush pad has been designed and analyzed to absorb the kinetic energy of an accidentally dropped spent nuclear fuel shipping cask into a 44 ft. deep cask unloading pool. Conventional analysis techniques available for evaluating a cask for tipping due to lateral seismic forces assume that the cask rests on a rigid surface. In this analysis, the cask (110 tons) sits on a stainless steel encased (0.25 in. top plate), polyurethane foam (4 ft. thick) crush pad. As the cask tends to rock due to horizontal seismic forces, the contact area between the cask and the crush pad is reduced, increasing the bearing stress, and causing the pivoting corner of the cask to depress into the crush pad. As the crush pad depresses under the cask corner, the pivot point shifts from the corner toward the cask center, which facilitates rocking and potential tipping of the cask. Subsequent rocking of the cask may deepen the depression, further contributing to the likelihood of cask tip over. However, as the depression is created, the crush pad is absorbing energy from the rocking cask. Potential tip over of the cask was evaluated by performing a non-linear, dynamic, finite element analysis with acceleration time history input. This time history analysis captured the effect of a deforming crush pad, and also eliminated conservatisms of the conventional approaches. For comparison purposes, this analysis was also performed with the cask sitting on a solid stainless steel crush pad. Results indicate that the conventional methods are quite conservative relative to the more exacting time history analysis. They also indicate that the rocking motion is less on the foam crush pad than on the solid stainless steel pad

  1. 77 FR 9591 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-02-17

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8... Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations by revising... Spent Fuel Storage Casks'' to include Amendment No. 8 to Certificate of Compliance (CoC) No....

  2. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Procedures for spent fuel storage cask submittals. 72.230... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel Storage Casks § 72.230 Procedures for spent fuel storage cask...

  3. 75 FR 33736 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Science.gov (United States)

    2010-06-15

    ... COMMISSION 10 CFR Part 72 RIN 3150--AI86 List of Approved Spent Fuel Storage Casks: MAGNASTOR System... Regulatory Commission (NRC) is proposing to amend its spent fuel storage cask regulations by revising the NAC International, Inc. (NAC), MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks''...

  4. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    Energy Technology Data Exchange (ETDEWEB)

    Samuel Bays; Ayodeji Alajo

    2010-05-01

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  5. Safety assessment of a metal cask under aircraft engine crash

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Hoon [Dept. of Mechanical and Automotive Engineering, Keimyung University, Daegu (Korea, Republic of); Choi, Woo Seok; Seo, Ki Seog [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    The structural integrity of a dual-purpose metal cask currently under development by the Korea Radioactive Waste Agency (KORAD) was evaluated, through numerical simulations and a model test, under high-speed missile impact reflecting targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from literature. In the impact scenario, a missile flying horizontally hits the top side of the cask, which is free standing on a concrete pad, with a velocity of 150 m/s. A simplified missile simulating a commercial aircraft engine was designed from an impact load-time function available in literature. In the analyses, the dynamic behavior of the metal cask and the integrity of the containment boundary were assessed. The simulation results were compared with the test results for a 1:3 scale model. Although the dynamic behavior of the cask in the model test did not match exactly with the prediction from the numerical simulation, other structural responses, such as the acceleration and strain history during the impact, showed very good agreement. Moreover, the containment function of the cask survived the missile impact as expected from the numerical simulation. Thus, the procedure and methodology adopted in the structural numerical analyses were successfully validated.

  6. Response of spent fuel transportation casks to explosive loadings

    International Nuclear Information System (INIS)

    Casks for the transportation of spent power reactor fuel can be exposed to explosive loadings from several causes. Exposure can come from an accident involving a propane or other hydrocarbon tanker, from an accident involving military or industrial explosives, or from deliberate sabotage. The regulations for the design of these casks do not specifically include requirements for resistance to blast loadings, but the hypothetical accident sequence that the casks are required to survive assure some measure of blast resistance. To perform accurate risk and security assessments, this blast resistance must be quantified. This paper will discuss the methodology used to determine the blast resistance of a representative rail and a representative truck spent fuel transportation cask. The methodology discussed in this paper can be used to determine the response to explosive loadings other than the one discussed in this paper or to determine the effect of explosive loadings on other casks. Due to the sensitive nature of this topic, this paper is intentionally vague on a number of parameters used in the analyses

  7. Safety Tests of Concrete Storage Cask for Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    In preparation for the timely installation of interim storage facility for spent nuclear fuel (SF), KORAD is developing domestic models of SF storage systems and the concrete storage cask is one of them. A concrete cask consists of a metallic canister which confines SF with welded closure and a concrete overpack which provides radiation shielding and physical protection to the canister. The safety requirements for a SF storage cask is well established in US and summarized in regulatory guides such as NUREG-1536. KAERI has been performing tests of the concrete cask to demonstrate its safety and compliance to the regulatory requirements with high priority stipulated in NUREG-1536. The test program includes the structural performance tests under tip-over and earthquake and decay heat removal test under normal, off-normal and accident conditions. In this paper, brief introduction to the structural tests and their results are provided. Safety tests to demonstrate the safety of KORAD21C concrete storage cask were successfully performed. The structural integrity during tip-over and earthquake were demonstrated with scale model tests and the results are analyzed in comparison with safety analysis results

  8. Effects of high burnup on spent-fuel casks

    International Nuclear Information System (INIS)

    Utility fuel managers have become very interested in higher burnup fuels as a means to reduce the impact of refueling outages. High-burnup fuels have significant effects on spent-fuel storage or transportation casks because additional heat rejection and shielding capabilities are required. Some existing transportation casks have useful margins that allow shipment of high-burnup fuel, especially the NLI-1/2 truck cask, which has been relicensed to carry pressurized water reactor (PWR) fuel with 56,000 MWd/ton U burnup at 450 days of cooling time. New cask designs should consider the effects of high burnup for future use, even though it is not commercially desirable to include currently unneeded capability. In conclusion, the increased heat and gamma radiation of high-burnup fuels can be accommodated by additional cooling time, but the increased neutron radiation source cannot be accommodated unless the balance of neutron and gamma contributions to the overall dose rate is properly chosen in the initial cask design. Criticality control of high-burnup fuels is possible with heavily poisoned baskets, but burnup credit in licensing is a much more direct means of demonstrating criticality safety

  9. Shielding calculations for spent CANDU fuel transport cask

    International Nuclear Information System (INIS)

    CANDU spent fuel discharged from the reactor core contains Pu, so, a special attention must be focussed into two directions: tracing for the fuel reactivity in order to prevent critical mass formation and personnel protection during the spent fuel manipulation. Shielding analyses, an essential component of the nuclear safety, take into account the difficulties occurred during the manipulation, transport and storage of spent fuel bundles, both for personnel protection and impact on the environment. The main objective here consists in estimations on radiation doses in order to reduce them under specified limit values. In order to perform the shielding calculations for the spent fuel transport cask three different codes were used: XSDOSE code and MORSE-SGC code, both incorporated in the SCALE4.4a system, and PELSHIE-3 code, respectively. As source of radiation one spent standard CANDU fuel bundle was used. All the geometrical and material data, related to the transport casks, were considered according to the shipping cask type B model, whose prototype has been realized and tested in the Institute for Nuclear Research Pitesti. The radial gamma dose rates estimated to the cask wall and in air, at different distances from the cask, are presented together with a comparison between the dose rates values obtained by all three recipes of shielding calculations. (authors)

  10. Vestibule and Cask Preparation Mechanical Handling Calculation

    Energy Technology Data Exchange (ETDEWEB)

    N. Ambre

    2004-05-26

    The scope of this document is to develop the size, operational envelopes, and major requirements of the equipment to be used in the vestibule, cask preparation area, and the crane maintenance area of the Fuel Handling Facility. This calculation is intended to support the License Application (LA) submittal of December 2004, in accordance with the directive given by DOE correspondence received on the 27th of January 2004 entitled: ''Authorization for Bechtel SAIC Company L.L.C. to Include a Bare Fuel Handling Facility and Increased Aging Capacity in the License Application, Contract Number DE-AC28-01R W12101'' (Ref. 167124). This correspondence was appended by further correspondence received on the 19th of February 2004 entitled: ''Technical Direction to Bechtel SAIC Company L.L. C. for Surface Facility Improvements, Contract Number DE-AC28-01R W12101; TDL No. 04-024'' (Ref. 16875 1). These documents give the authorization for a Fuel Handling Facility to be included in the baseline. The limitations of this preliminary calculation lie within the assumptions of section 5 , as this calculation is part of an evolutionary design process.

  11. Final version dry cask storage study

    International Nuclear Information System (INIS)

    This report was prepared in response to Section 5064 of the Nuclear Waste Policy Amendments Act of 1987 (the Amendments Act--Public Law 100-203), which directs the Secretary of Energy to conduct a study of the use of dry-cask-storage technology for storing spent fuel at the sites of civilian nuclear reactors until a geologic repository is available. In conducting this study, whose results are being reported to the Congress, the Secretary was to consider such factors as costs, effects on human health and the environment, and the extent to which the Nuclear Waste Fund can and should be used to provide funds for at-reactor storage. In addition, the Secretary was to consult with the Nuclear Regulatory Commission (NRC), include NRC comments in the report, and solicit the views of State and local governments and the public. The study performed in response to these requirements was based largely on data published by the DOE or the NRC or included in documents issued by the DOE. Among the DOE documents are the 1987 MRS proposal to the Congress and a subsequent report, prepared to supply the Congress with additional information on the MRS facility. Because in evaluating dry storage at reactor sites it is necessary to take into account other options for meeting storage needs, this study covered all forms of at-reactor storage. 107 refs., 15 figs., 10 tabs

  12. Initial version, dry cask storage study

    International Nuclear Information System (INIS)

    This report was prepared to study the use of dry cask storage for storing spent fuel at the sites of civilian nuclear reactors until a geologic repository is available. In conducting this study, whose results are to be reported to the Congress, the Secretary is to consider such factors as costs, effects on human health and the environment, and the extent to which the Nuclear Waste Fund can and should be used to provide funds for at-reactor storage. In addition, the Secretary is to consult with the Nuclear Regulatory Commission (NRC), include NRC comments in the report, and solicit the views of State and local governments and the public. The study performed in response to these requirements was based largely on data published by the NRC or included in documents issued by the DOE. Among the DOE documents are the MRS proposal to the Congress and a subsequent report, prepared to supply the Congress with additional information on the MRS facility. Because in evaluating dry storage at reactor sites it is necessary to take into account other options for meeting storage needs, this study covered all forms of at-reactor storage. 85 refs., 5 figs, 12 tabs

  13. Production of casks acceptable for final storage by subsequent treatment of prefilled casks

    International Nuclear Information System (INIS)

    During the operation and the decommissioning of nuclear facilities also radioactive waste material which cannot be encompassed under the general standard waste categories arises. To transfer these types of waste material to interim/final repositories a conditioning/treatment is necessary in most cases. The acceptance conditions of the interim and final repositories require a conditioning considering the type of waste, the specific activities, and the casks to be used. A possible way of conditioning e. g. liquid waste (resins, filter aid, etc.) is to fill the waste into thick-wall casks, if necessary with additional shielding and subsequent drying res. draining. This presentation shall show the experiences and the results gained from the conditioning of these types of middle and higher activated waste. In the NPP Neckar (GKN) 14 ea. 200-I-rolling hoop drums and in the NPP Brokdorf (KBR) 83 ea. mouldings filled with granular resins were stored. 32 200-I-drums with higher activated filters, sludge, as well as mixed waste were located in shielded areas of the drum storage. (orig.)

  14. Standard review plan for dry cask storage systems. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

    The Standard Review Plan (SRP) For Dry Cask Storage Systems provides guidance to the Nuclear Regulatory Commission staff in the Spent Fuel Project Office for performing safety reviews of dry cask storage systems. The SRP is intended to ensure the quality and uniformity of the staff reviews, present a basis for the review scope, and clarification of the regulatory requirements. Part 72, Subpart B generally specifies the information needed in a license application for the independent storage of spent nuclear fuel and high level radioactive waste. Regulatory Guide 3.61 {open_quotes}Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask{close_quotes} contains an outline of the specific information required by the staff. The SRP is divided into 14 sections which reflect the standard application format. Regulatory requirements, staff positions, industry codes and standards, acceptance criteria, and other information are discussed.

  15. Contract Report for Usage Inspection of KN-12 Transport Cask

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. C.; Seo, K. S.; Bang, K. S.; Cho, I. J.; Kim, D. H.; Min, D. K

    2007-03-15

    The usage inspection of the KN-12 spent nuclear fuel transport package was performed to receive the license for reuse. According to the Korea Atomic Energy Act, all type B transport package should receive and pass the usage inspection every five years. The KN-12 transport cask was designed to transport twelve spent PWR fuel assemblies under wet and dry conditions. The cask was developed and licensed in 2002 in accordance with the Korea and the IAEA's safe transport regulations. The areas of usage inspection include: visual inspection, nondestructive weld inspection, load test, maximum operating pressure test, leakage test, shielding test, thermal test, external surface contamination test. In the results of the usage inspection, the damage or defect could not found out and the performance of the cask was maintained according to the requirements of the regulation. Therefore, the usage inspection was successfully performed to acquire the license for the reuse.

  16. CONTAINMENT EVALUATION OF BREACHED AL-SNF FOR CASK TRANSPORT

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D. W.; Sindelar, R. L.; Iyer, N. C.

    2005-11-07

    Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site. To enter the U.S., the cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Al-SNF is subject to corrosion degradation in water storage, and many of the fuel assemblies are ''failed'' or have through-clad damage. A methodology has been developed with technical bases to show that Al-SNF with cladding breaches can be directly transported in standard casks and maintained within the allowable release rates. The approach to evaluate the limiting allowable leakage rate, L{sub R}, for a cask with breached Al-SNF for comparison to its test leakage rate could be extended to other nuclear material systems. The approach for containment analysis of Al-SNF follows calculations for commercial spent fuel as provided in NUREG/CR-6487 that adopts ANSI N14.5 as a methodology for containment analysis. The material-specific features and characteristics of damaged Al-SNF (fuel materials, fabrication techniques, microstructure, radionuclide inventory, and vapor corrosion rates) that were derived from literature sources and/or developed in laboratory testing are applied to generate the four containment source terms that yield four separate cask cavity activity densities; namely, those from fines; gaseous fission product species; volatile fission product species; and fuel assembly crud. The activity values, A{sub 2}, are developed per the guidance of 10CFR71. The analysis is performed parametrically to evaluate maximum number of breached assemblies and exposed fuel area for a proposed shipment in a cask with a test leakage rate.

  17. Dry Spent Fuel Cask Transporter equipment design, testing, and operational features

    International Nuclear Information System (INIS)

    The United States Department of Energy (DOE) has established a program for the testing of a variety of dry spent fuel storage casks. The program is being conducted at the Idaho National Engineering Laboratory (INEL) by EG and G Idaho Inc. Testing of storage casks at INEL requires that large storage casks (max. gross wt. 127.1 Mg) be moved and positioned from/to an indoor loading location to an outdoor storage pad. A Dry Spent Fuel Cask Transporter has been developed to safely, conveniently, and economically transport/handle a variety of storage casks within and around the confines of nuclear sites and facility

  18. Thermoelectric Powered Wireless Sensors for Dry-Cask Storage

    Science.gov (United States)

    Carstens, Thomas Alan

    This study focuses on the development of self-powered wireless sensors. These sensors can be used to measure key parameters in extreme environments; e.g., temperature monitoring for spent nuclear fuel during dry-cask storage. This study has developed a design methodology for these self-powered monitoring systems. The main elements that constitute this work consist of selecting and testing a power source for the wireless sensor, determination of the attenuation of the wireless signal, and testing the wireless sensor circuitry in an extreme environment. OrigenArp determined the decay heat and gamma/neutron source strength of the spent fuel throughout the service life of the dry-cask. A first principles analysis modeled the temperatures inside the dry-cask. A finite-element heat transfer code calculated the temperature distribution of the thermoelectric and heat sink. The temperature distributions determine the power produced by the thermoelectric. It was experimentally verified that a thermoelectric generator (HZ-14) with a DC/DC converter (Linear Technology LTC3108EDE) can power a transceiver (EmbedRF) at condition which represent prototypical conditions throughout and beyond the service life of the dry-cask. The wireless sensor is required to broadcast with enough power to overcome the attenuation from the dry-cask. It will be important to minimize the attenuation of the signal in order to broadcast with a small transmission power. To investigate the signal transmission through the dry-cask, CST Microwave Studio was used to determine the scattering parameter S2,1 for a horizontal dry-cask. Important parameters that can influence the transmission of the signal are antenna orientation, antenna placement, and transmission frequency. The thermoelectric generator, DC/DC converter, and transceiver were exposed to 60Co gamma radiation (exposure rate170.3 Rad/min) at the University of Wisconsin Medical Radiation Research Center. The effects of gamma radiation on the

  19. Quality assurance in a cask fleet parts control system

    International Nuclear Information System (INIS)

    This paper discusses applicable portions of the eighteen Quality Assurance criteria of Subpart H, 10 CFR 71 which are incorporated into a relational data base system which has been designed to manage the spare parts control system for a fleet of spent nuclear fuel casks. The system includes not only parts in warehouse storage but parts in use in the field plus casks, ancillary equipment, test equipment, support devices, and even personnel. It provides a high degree of assurance that any device for which a condition for certification has expired will be flagged for recertification testing or removal from service well before the critical date

  20. Spent fuel storage cask testing and operational experience at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Spent-fuel storage cask research, development, and demonstration activities are being performed for the U.S. Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) as a part of the storage cask testing program. The cask testing program at federal sites and other locations supports the Nuclear Waste Policy Act (NWPA) and DOE objectives for cooperative demonstrations with the cask vendors and utilities for development of at-reactor dry cask storage capabilities for spent nuclear fuel assemblies. One research and development program for the storage cask performance testing of metal storage cask was initiated through a cooperative agreement between Virginia Power and DOE in 1984. The performance testing was conducted for the DOE and the Electric Power Research Institute by the Pacific Northwest laboratory, operated for DOE by Battelle Memorial Institute, and the Idaho National Engineering Laboratory (INEL), operated for DOE by EG ampersand G Idaho, Inc. In 1988 a cooperative agreement was entered into by DOE with Pacific Sierra Nuclear Associates (PSN) for performance testing of the PSN concrete Ventilated Storage Cask. Another closely related activity involving INEL is a transportable storage cask project identified as the Nuclear Fuel Services Spent-Fuel Shipping/Storage Cask Demonstration Project. The purpose of this project is to demonstrate the feasibility of packing, transporting, and storing commercial spent fuel in dual-purpose transport/storage casks

  1. Computational fluid dynamics analysis for K24B cask design with burnup credit

    International Nuclear Information System (INIS)

    Korea Nuclear Engineering Service Corp. (KONES) has designed K24B cask for the storage and the transportation of 24 (CE-type 16x16) PWR assemblies. K24B cask is designed with considering burnup credit of spent fuel. In order to remove heat from the fuel assemblies effectively, the flow channels in the upper and the lower part of fuel assemblies are set up to promote the natural convection. Computational fluid dynamics analysis is carried out to estimate and assure the thermal integrity of K24B cask. Conduction and radiation heat transfer through the cask components and the natural convective heat transfer in the cask are simulated. As a result of the analysis, the maximum temperatures of the cask components are maintained below the operating temperature for the safety. Therefore, the design of K24B cask can satisfy the safety limit. (author)

  2. The interim storage facility with dry storage casks and its safeguards activity

    International Nuclear Information System (INIS)

    Recyclable-Fuel Storage Company (RFS) is constructing an interim storage facility of spent fuel at Recyclable-Fuel Storage Center (RFSC) in Aomori Prefecture. Metallic dry casks are employed to contain the spent fuel from nuclear power plants and to serve for about 50 years in RFSC. Metallic dry casks have already been used for dry cask storage facility at Tokai No.2 power station of Japan Atomic Power Company. But, RFSC is not exactly the same as the dry cask storage facility at Tokai No.2 power station, for example, cask transportation between facilities and no hot cells. Therefore, additional safeguards activities are necessary. The outline of the design and handling of metallic dry casks at RFSC and the currently developing status of safeguards activity such as containment and surveillance for the cask receipt and storage at RFSC, etc are described. (author)

  3. Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

    CERN Document Server

    Poulson, D; Guardincerri, E; Morris, C L; Bacon, J D; Plaud-Ramos, K; Morley, D; Hecht, A

    2016-01-01

    Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, the cask contents can be confirmed with high confidence in less than two days exposure. Similar results can be obtained by moving a smaller detector to view the cask from multiple angles.

  4. IMPACT ANALYSIS OF SPENT FUEL DRY CASKS UNDER ACCIDENTAL DROP SCENARIOS

    International Nuclear Information System (INIS)

    A series of analyses were performed to assess the structural response of spent nuclear fuel dry casks subjected to various handling and on-site transfer events. The results of these analyses are being used by the Nuclear Regulatory Commission (NRC) to perform a probabilistic risk assessment (PRA). Although the PRA study is being performed for a specific nuclear plant, the PRA study is also intended to provide a framework for a general methodology that could also be applied to other dry cask systems at other nuclear plants. The dry cask system consists of a transfer cask, used for handling and moving the multi-purpose canister OLIIpC that contains the fuel, and a storage cask, used to store the MPC and fuel on a concrete pad at the site. This paper describes the analyses of the casks for two loading events. The first loading consists of dropping the transfer cask while it is lowered by a crane to a concrete floor at ground elevation. The second loading consists of dropping the storage cask while it is being transferred to the concrete storage pad outdoors. Three dimensional finite element models of the transfer cask and storage cask, containing the MPC and fuel, were utilized to perform the drop analyses. These models were combined with finite element models of the target structures being impacted. The transfer cask drop analyses considered various drop heights for the cask impacting the reinforced concrete floor at ground level. The finite element model of the target included a section of the concrete floor and concrete wall supporting the floor. The storage cask drop analyses evaluated a 30.5 cm (12 in.) drop of the cask impacting three different surfaces: reinforced concrete, asphalt, and gravel

  5. Logistics management for storing multiple cask plug and remote handling systems in ITER

    International Nuclear Information System (INIS)

    Highlights: ► We model the logistics management problem in ITER, taking into account casks of multiple typologies. ► We propose a method to determine the best position of the casks inside a given storage area. ► Our method obtains the sequence of operations required to retrieve or store an arbitrary cask, given its storage place. ► We illustrate our method with simulation results in an example scenario. -- Abstract: During operation, maintenance inside the reactor building at ITER (International Thermonuclear Experimental Reactor) has to be performed by remote handling, due to the presence of activated materials. Maintenance operations involve the transportation and storage of large, heavyweight casks from and to the tokamak building. The transportation is carried out by autonomous vehicles that lift and move beneath these casks. The storage of these casks face several challenges, since (1) the cask storage area is limited in space, and (2) all casks have to be accessible for transportation by the vehicles. In particular, casks in the storage area may block other casks, so that the former has to be moved to a temporary position to give way to the latter. This paper addresses the challenge of managing the logistics of cask storage, where casks may have different typologies. In particular, we propose an approach to (1) determine the best position of the casks inside the storage area, and to (2) obtain the sequence of operations required to retrieve and store an arbitrary cask from/to a given storage place. A combinatorial optimization approach is used to obtain solutions to both these problems. Simulation results illustrate the application of the proposed method to a simple scenario

  6. Studies and research concerning BNFP: advanced cask handling studies

    International Nuclear Information System (INIS)

    Cask turnaround times at loading and unloading sites can be improved by providing better working conditions, improved safety, reduced decontamination time, training, and where practical to do so, improved facility design. This report consists of treatments of several of these topics with the common goal of improving operational efficiency

  7. Separator assembly for use in spent nuclear fuel shipping cask

    Science.gov (United States)

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  8. Interim Dry Storage of Spent Fuel in Casks

    International Nuclear Information System (INIS)

    French option for the back end of the fuel cycle is reprocessing of used fuel and recycling the fissile material, except some very specific fuel stored in vaults (dry conditions). Used fuel management solutions studied by AREVA for various countries allow for either direct transport to the reprocessing plant, or interim storage and transport after storage of used fuel. Interim storage solutions are wet storage or dry storage (DSC, metal casks or vault systems). When the decision on used fuel management has been postponed, some extension of interim storage duration is considered, therefore it becomes necessary to study used fuel and cask material behaviour and deterioration mechanisms. One objective of this R&D was to review research efforts on spent fuel behaviour and Dry storage experience in casks. Particularly we were interested in the assessment of retrievability of fuel after storage for further use. A review therefore, was made of the effect of storage time/ temperatures and of loading/ drying operation on used fuel integrity. R&D programmes were also carried out on the evaluation of cask materials in long term, especially materials susceptible to degradation

  9. Implementation of response function concept for spent fuel cask analyses

    International Nuclear Information System (INIS)

    Due to the uncertain schedule about the first disposal of the large quantity of spent nuclear fuel (SNF) accumulated at the US commercial nuclear power plants, and due to the wide range of burnups and cooling times of the SNF, it is urgent to develop a quick and realistic method for analyzing an interim-storage or shipping package of SNF. The existing method uses design-basis SNF, and it is unnecessarily conservative and therefore uneconomic. This paper demonstrates the use of response-function concept for shielding and criticality analysis for a commercial SNF shipping cask. A PC-based computer code is written for this purpose. The program allows users to perform accurate shielding and criticality analyses for any selected cask payload on real-time basis. The results are less conservative, but more realistic than that of the design-basis analyses. One must be noted, however, that the response function is cask-specific. Therefore, the concept is most beneficial to the major cask type which is to be repeatedly used for a large number of SNF shipments

  10. Structural analysis of closure bolts for shipping casks

    Energy Technology Data Exchange (ETDEWEB)

    Mok, G.C.; Fischer, L.E.

    1993-04-01

    This paper identifies the active forces and moments in a closure bolt of a shipping cask. It examines the interactions of these forces/moments and suggest simplified methods for their analysis. The paper also evaluates the role that the forces and moments play in the structure integrity of the closure bolt and recommends stress limits and desirable practices to ensure its integrity.

  11. Development of a dry storage cask for PWR spent fuel

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power Co., Ltd.(KHNP), which operates all the nuclear power plants in Korea, is developing a new dry storage cask to store twenty four spent fuel assemblies generated from pressurized water reactors for at-reactor or away-from-reactor interim storage facility in Korea. The dry storage cask is designed and evaluated according to the requirements of the IAEA, the US NRC and the Korean regulations for the dry spent fuel storage system. It provides confinement, radiation shielding, structural integrity, subcritical control and passive heat removal for normal and accident conditions. The dry storage cask consists of a dual purpose canister providing a confinement boundary for the PWR spent fuel, and a storage overpack providing a structural and radiological boundary for long-term storage of the canister placed inside it. The overpack is constructed by a combination of steel and concrete, and is equipped with penetrating ducts near its lower and upper extremities to permit natural circulation of air to provide for the passive cooling of the canister and the contained spent fuel assemblies. This paper describes development status, description, design criteria, evaluation and demonstration tests of the dry storage cask. (authors)

  12. Monitored Retrievable Storage conceptual system study: cask-in-trench

    International Nuclear Information System (INIS)

    This report provides a description of the Cask-in-Trench Storage Concept which meets a specified set of requirements; an estimate of the costs of construction, operation and decommissioning of the concept; the costs required to expand the facility throughput and storage capability; and the life cycle costs of the facility. 22 figures, 34 tables

  13. Monte Carlo shipping cask calculations using an automated biasing procedure

    International Nuclear Information System (INIS)

    This paper describes an automated biasing procedure for Monte Carlo shipping cask calculations within the SCALE system - a modular code system for Standardized Computer Analysis for Licensing Evaluation. The SCALE system was conceived and funded by the US Nuclear Regulatory Commission to satisfy a strong need for performing standardized criticality, shielding, and heat transfer analyses of nuclear systems

  14. Design of casks: incorporating operational feedback from maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Bimet, F.; Hartenstein, M. [COGEMA Logistics, Saint Quentin (France)

    2004-07-01

    Casks are designed to conform to regulations and to client specifications. Essential areas such as easy operation, low costs of maintenance, low operation and maintenance doses, limited waste, are not explicitly covered. Notwithstanding, COGEMA LOGISTICS uses all feedback available, so that casks are designed to be easy, safe and economical to operate and maintain. Maintenance is an activity where you do spot items that old-time designers could have made better, and things that users should not have done. Thanks to quality assurance, there are a number of data available, waiting to be collected and exploited; they have to be identified, located, retrieved, and analysed. Using information such as wear, damage, use of spare parts, access problems helps to make casks ever better. It leads to more efficient concepts, and to upgrades on existing designs; it also allows to integrate environmental considerations, inter alia in the choice of materials and in maintenance methods. It is necessary for the designer to interact with the users, the cask owners, the maintenance providers, in order to gather all relevant information and events. This is made easier when all these actors are ''under one roof'', or have very close ties. This paper presents COGEMA LOGISTICS methods for collecting and analysing all these experiences waiting to be used. It explains our process for analysing data, preparing yearly reports that are made available to our designers. It describes how each new design is subject to a maintainability study, using this feedback, so that the cask safety is always assured, that radiological doses are kept to a minimum, and that operating and maintenance costs will remain as low as possible.

  15. Design of casks: incorporating operational feedback from maintenance

    International Nuclear Information System (INIS)

    Casks are designed to conform to regulations and to client specifications. Essential areas such as easy operation, low costs of maintenance, low operation and maintenance doses, limited waste, are not explicitly covered. Notwithstanding, COGEMA LOGISTICS uses all feedback available, so that casks are designed to be easy, safe and economical to operate and maintain. Maintenance is an activity where you do spot items that old-time designers could have made better, and things that users should not have done. Thanks to quality assurance, there are a number of data available, waiting to be collected and exploited; they have to be identified, located, retrieved, and analysed. Using information such as wear, damage, use of spare parts, access problems helps to make casks ever better. It leads to more efficient concepts, and to upgrades on existing designs; it also allows to integrate environmental considerations, inter alia in the choice of materials and in maintenance methods. It is necessary for the designer to interact with the users, the cask owners, the maintenance providers, in order to gather all relevant information and events. This is made easier when all these actors are ''under one roof'', or have very close ties. This paper presents COGEMA LOGISTICS methods for collecting and analysing all these experiences waiting to be used. It explains our process for analysing data, preparing yearly reports that are made available to our designers. It describes how each new design is subject to a maintainability study, using this feedback, so that the cask safety is always assured, that radiological doses are kept to a minimum, and that operating and maintenance costs will remain as low as possible

  16. Heat transfer investigations within dry spent fuel casks

    International Nuclear Information System (INIS)

    For studying the heat transfer processes and predicting the maximum spent fuel element surface temperature in a spent fuel assembly (SFA) transported in a dry cask, model experiments have been performed with a gas-filled model cask containing a simplified electrically heated model of a WWER-type SFA with 90 fuel elements. The temperature distribution of the SFA model is measured for different heat rates under vacuum in the model cask, and under normal pressure and overpressure (0.1 ... 0.7 MPa) for several cooling gases (air, argon, helium) in order to separately investigate heat transfer processes by radiation and convection/conduction. The measuring results were compared with the calculations. Computer programmes as well as simplified calculation methods for temperature prediction were developed and checked. The results obtained are also useful for thermal analyses in the field of the dry storage of SFAs in a cask or can. Specifically it was found that: The heat removal from the SFA can be considerably improved by increasing the internal cask pressure or by using helium as coolant. The radiant heat exchange in the SFA model can be calculated with sufficient accuracy by means of a computer programme developed in 1978 or by means of a simplified analytical representation shown in the final report. Both methods are directly applicable to the original SFA and useful in order to approximately calculate the maximum SFE surface temperature under normal pressure, if the fraction of heat transferred by radiation is allowed for. For the calculation of the total heat transfer a computer programme was developed and verified, which completely permits the temperature prediction of the SFA model in dependence on heat rate, type of gaseous coolant and coolant pressure. This computer programme can be directly applied to the original SFA for the calculation of the maximum SFE surface temperature

  17. Opportunities to increase the productivity of spent fuel shipping casks in the United States

    International Nuclear Information System (INIS)

    Trends indicate that future transportation requirements for spent fuel will be different from those anticipated when the current generation of casks and vehicles was designed. Increased storage capacity at most reactors will increase the average post irradiation age of the spent fuel to be transported. A scenario is presented which shows the 18 casks currently available should be sufficient until approximately 1983. Beyond this time, it appears that an adequate transportation system can be maintained by acquiring, as needed, casks of current designs and new casks currently under development. Spent fuel transportation requirements in the post-1990 period can be met by a new generation of casks specifically designed to transport long-cooled fuel. In terms of the number of casks needed, productivity may be increased by 19% if rail cask turnaround time is reduced to 4 days from the current range of 6.5 to 8.5 days. Productivity defined as payloads per cask year could be increased 62% if the turnaround time for legal weight truck casks were reduced from 12 hours to 4 hours. On a similar basis, overweight truck casks show a 28% increase in productivity

  18. Scoping study of casks shipped from the MRS facility to various repository sites

    International Nuclear Information System (INIS)

    The objective of this study was to determine the maximum number of specialized repository waste packages that could be shipped from the Monitored Retrievable Storage (MRS) facility in Pb-, Fe-, and U-shielded casks weighing 200,000 or 300,000 lbs. The study included 18 different waste packages designed for the Salt, Tuff, and Basalt repositories. Nine of these contained consolidated PWR fuel pins, and nine contained consolidated BWR fuel pins. Discrete ordinates calculations were performed to determine the neutron and gamma shield thicknesses that would ensure a dose rate of 10 millirem/hr, 10 ft from the centerline of the cask(s). Over 100 casks of particular interest have been identified, while preliminary design information is also given for 522 casks of potential interest. Relative to the 200,000-lb casks, 50 to 100% more fuel may be shipped in the larger 300,000-lb casks. Placing the spent fuel canisters in overpacks prior to shipment from the MRS will reduce the net payload by 30 to 50%. The highest-capacity cask/waste package combination studied corresponds to a 300,000-lb U-shielded cask containing 84 consolidated PWR fuel assemblies in 21 canisters, or 171 consolidated BWR fuel assemblies in 19 canisters. Criticality analyses have shown these high-capacity casks to be safely subcritical - even if all the canisters were loaded with unirradiated LWR fuel containing 3.4 wt % U-235

  19. Research Reactor Spent Fuel Transfer/Storage Cask with Application to TRIGA Fuel - Designed Cask Shielding Independent Evaluation

    International Nuclear Information System (INIS)

    Institute for Nuclear Research (INR) Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies (dual-core concept involves independent operation of TRIGA 14 MW Steady-State Reactor and TRIGA Annular-Core Pulsing Reactor at each end of a large pool). In May 2006, TRIGA 14 MW SSR core was fully converted to Low Enriched Uranium (LEU 20 wt% 235U) fuel, according to Reduced Enrichment for Research and Test Reactors agreements and current worldwide non-proliferation efforts. Paper presents a shielding independent evaluation applied to designed transfer/ storage cask for TRIGA INR spent fuel, a mandatory step in preparation of the documentation required for spent fuel transfer/storage cask authorisation process. Fuel elements irradiation was modelled by assuming constant power for entire residence time inside reactor core, for 14 MW reactor operation power and two different scenarios characteristic for accident calculations according to TRIGA 14 MW SSR safety report and reactor operation experience. The discharged spent LEU fuel was cooled down for 2 and 5 years, respectively. Source term assessment and spent fuel characteristic parameters estimation were done by means of ORIGEN-S burn-up code (included in Oak Ridge National Laboratory's SCALE6 package) with specific cross-sections libraries, updating data for each burn-up step. For the transfer/storage cask shielding analysis, two different cases have been considered, the main difference residing in TRIGA fuel elements loading. The radiation dose rates to the transfer/storage cask wall and in air at different distances from the cask have been estimated by means of MAVRIC/Monaco shielding 3D Monte Carlo code included in ORNL's SCALE6 package. (author)

  20. Aplikasi Analisis DNA dalam Bidang Forensik

    Directory of Open Access Journals (Sweden)

    Elza Ibrahim Auerkari

    2015-10-01

    Full Text Available Analisis DNA dalam bidang forensik merupakan teknik yang relatif baru dan berkembang pesar sesuai dengan peningkatan kualitas dan kuantitas kriminalitas disamping dapat digunakan dalam penentuan hubungan keluarga. Permasalahannya adalah bagaimana kemampuan analisis DNA ini dalam mengidentifikasi individu pada kasus-kasus tersebut. Dari 3,3 milyar pasang basa yang membentuk genom manusia, terdapat sekitar 3 juta perbedaan di antara setiap dua individu. Untuk tujuan identifikasi DNA dalam bidang forensik, regio yang sangat penting adalah lokus polimorfik DNA termasuk regio ukuran satelit (Satelite sequence pada bagian yang tidak mengkode produk tertentu dari genom manusia. Bila frekuensi folimorfis DNA pada suatu populasi diketahui, probabilitas dari identifikasi, lokus polimorfik dengan frekuensi yang diketahui dalam suatu populasi dapat dipilih sebagai DNA maeker. Analisis DNA merupakan suatu metode yang sangat potensial yang dewasa ini telah diterima secara luas sebagai suatu cara identifikasi dalam bidang forensik, sebab hanya dibutuhkan sedikit sampel saja yang dapat diambil dari semua sel berinti di seluruh tubuh. Penggunaan analisis DNA dan bank data DNA berkembang dengan pesat serta merupakan sarana yang penting sebagai pelengkap terhadap bidang kedokteran dan kedokteran gigi forensik lainnya. Guna efisiensi identifikasi di bidang forensik dianjurkan agar metode-metode yang ada dikombinasi.

  1. PRINSIP DAN METODE ANALISIS RISIKO KESEHATAN LINGKUNGAN

    Directory of Open Access Journals (Sweden)

    Defriman Djafri

    2014-04-01

    Full Text Available Analisis risiko kesehatan lingkungan merupakan penilaian atau penaksiran risiko kesehatan yang bisa terjadi di suatu waktu pada populasi manusia berisiko. Kajian prediktif ini menghasilkan karakteristik risiko secara kuantitatif, pilihan-pilihan manajemen risiko dan strategi komunikasi untuk meminimalkan risiko tersebut. Data kualitas lingkungan yang bersifat agent specific dan site specific, karakteristik antropometri dan pola aktivitas populasi terpajan dibutuhkan untuk kajian ini.

  2. PRINSIP DAN METODE ANALISIS RISIKO KESEHATAN LINGKUNGAN

    OpenAIRE

    Defriman Djafri

    2014-01-01

    Analisis risiko kesehatan lingkungan merupakan penilaian atau penaksiran risiko kesehatan yang bisa terjadi di suatu waktu pada populasi manusia berisiko. Kajian prediktif ini menghasilkan karakteristik risiko secara kuantitatif, pilihan-pilihan manajemen risiko dan strategi komunikasi untuk meminimalkan risiko tersebut. Data kualitas lingkungan yang bersifat agent specific dan site specific, karakteristik antropometri dan pola aktivitas populasi terpajan dibutuhkan untuk kajian ini.

  3. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP)

  4. Size and transportation capabilities of the existing U.S. cask fleet

    International Nuclear Information System (INIS)

    This paper investigates the current spent nuclear fuel cask fleet capability in the United States. It assesses the degree to which the current fleet would be available, as a contingency, until proposed Office of Civilian Radioactive Waste Management casks become operational. A limited fleet of ten spent fuel transportation casks is found to be readily available for use in Federal waste management efforts over the next decade

  5. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    Science.gov (United States)

    Muhamad, Shalina Sheik; Hamzah, Mohd Arif Arif B.

    2014-02-01

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP).

  6. Design analysis report for the TN-WHC cask and transportation system

    Energy Technology Data Exchange (ETDEWEB)

    Brisbin, S.A., Fluor Daniel Hanford

    1997-02-13

    This document presents the evaluation of the Spent Nuclear Fuel Cask and Transportation System. The system design was developed by Transnuclear, Inc. and its team members NAC International, Nelson Manufacturing, Precision Components Corporation, and Numatec, Inc. The cask is designated the TN-WHC cask. This report describes the design features and presents preliminary analyses performed to size critical dimensions of the system while meeting the requirements of the performance specification.

  7. Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

    OpenAIRE

    Poulson, D.; Durham, J. M.; Guardincerri, E.; Morris, C. L.; Bacon, J. D.; Plaud-Ramos, K.; Morley, D.; Hecht, A.

    2016-01-01

    Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first applicati...

  8. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    Energy Technology Data Exchange (ETDEWEB)

    Muhamad, Shalina Sheik [Prototype and Plant Development Center, Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000, Kajang, Selangor (Malaysia); Hamzah, Mohd Arif Arif B. [Prototype and Plant Development Center, Technical Support Division Malaysian Nuclear Agency, Bangi, 43000, Kajang, Selangor (Malaysia)

    2014-02-12

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP)

  9. Castor transport and storage casks for VVER and RBMK fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Gartz, R.; Gobler, A.; John, R.; Diersch, R. [GNB Gesellschaft fur Nuklear-Behalter mbH, Essen (Germany); Nemec, P. [Skoda Nuclear Machinery Plzen (Czech Republic)

    1998-12-31

    CASTOR casks have been successfully developed, manufactured and delivered for Russian type reactor fuel assemblies. These casks fulfill both the requirements for type B packages according to IAEA regulations and the requirements covering different accident situations to be assumed at the storage site. In the following, the CASTOR casks CASTOR 440/84, CASTOR RBMK and CASTOR VVER 1000 are described, the nuclear content is characterized and an overview about the status of licensing, manufacturing and delivery is given. (authors) 3 refs.

  10. Analisis SWOT pada Toko Lestari Rattan and Furniture Jl. Gatot Subroto No. 457 Medan

    OpenAIRE

    Desira, Sheila

    2011-01-01

    Penelitian ini bertujuan untuk menganalisis SWOT yang ada pada Toko Lestari Rattan and Furniture. Analisis SWOT merupakan penilaian lingkungan internal berupa kekuatan (strength) dan kelemahan (weakness) serta lingkungan eksternal berupa peluang (opportunity) dan ancaman (treath). Analisis SWOT akan menghasilkan strategi yang tepat bagi perusahaan. Analisis SWOT umumnya digunakan pada perusahaan besar. Penelitian ini melihat penerapan analisis SWOT pada UKM (Usaha Kecil Mene...

  11. The NINO [No Inspector, No Operator system] cask-loading safeguards system

    International Nuclear Information System (INIS)

    It is, in general difficult to determine by means of camera-surveillance techniques what is loaded into spent-fuel casks being prepared for shipment from light-water reactors to other reactors, reprocessing facilities, or long-term storage. Furthermore, the expected high frequency of cask loadings in the coming years would place too great a burden on the IAEA and Euratom inspectorates if each had to be observed by an inspector. For the case of shipment to other reactors and reprocessing facilities, the casks are soon opened and, in principle, their contents could be ascertained by direct inspection. In the case of long-term-storage facilities, the casks would stay sealed for years, thereby requiring the IAEA to know positively how many spent-fuel assemblies were loaded at the reactor and to have a continuity of knowledge of the cask's contents. It has been proposed instead that the facility operator place the cask seal on the cask within the field of view of a surveillance system linked to the cask seal. This solution, however, may not provide enough credibility for acceptance by the safeguards community. This paper presents an alternative to both inspector presence at cask loading and operator assistance in applying seals; this alternative is called the No Inspector, No Operator system (NINO)

  12. Shielding analysis of dual purpose casks for spent nuclear fuel under normal storage conditions

    International Nuclear Information System (INIS)

    Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, UO2 pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a 2 x 10 cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the 2 x 10 cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the 2 x

  13. Numerical simulation of ambient flow and thermal distributions in a spent fuel storage cask array

    Energy Technology Data Exchange (ETDEWEB)

    Michener, T. [Pacific Northwest National Laboratory, Richland WA (United States); Trent, D.S.; Guttmann, J.; Bajwa, C. [United States Nuclear Regulatory Commission, One White Flin North, Rockville MD (United States)

    2001-07-01

    At the request of the U.S. Nuclear Regulatory Commission (USNRC), the staff at the Pacific Northwest National Laboratory (PNNL) analyzed the thermal performance of the Utah Private Fuel Storage (PFS) using the TEMPEST computational fluid dynamics software. A three-dimensional section of the PFS with a total of 20 casks was modeled to estimate the ambient flow and temperature distributions surrounding the casks. The purpose of this analysis was to compute the cask inlet vent air temperature to be used for boundary conditions in a detailed analysis of an individual Holtec Hi-Storm 100 cask using the COBRA-SFS (Spent Fuel Storage) thermal hydraulic computer software. (author)

  14. Safety analysis report for packaging: the ORNL in-pile capsule shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Haynie, C.B.; Crowley, W.K.; Just, R.A.

    1977-11-01

    The ORNL in-pile capsule shipping cask is used to transport irradiated experimental capsules and spent fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures were used to determine the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations.

  15. STABILITY EVALUATION OF METAL CASK ATTACHED TO A TRANSFER PALLET DURING LONG-PERIOD SEISMIC MOTIONS

    Science.gov (United States)

    Kawaguchi, Shohei; Shirai, Koji; Kanazawa, Kenji

    Rocking behavior of unfixed body is affected by center of mass, material coefficient of restitution and so on. 2/5 scale metal cask model considering these parameter was used for seismic test to evaluate stability of grounding metal cask attached to a transfer pallet under the influence of long-period earthquake motion. The newest knowledge from seismic test indicates seismic motion with high velocity over 100 kine not always cause the raise of response velocity of metal cask because of energy consumption by cask sliding and impact deformation of concrete. And new estimation method (called "Window energy spectrum method") of earthquake response spectrum gives suitable evaluation of response energy.

  16. Experimental studies of free-standing spent fuel storage cask subjected to strong earthquakes

    International Nuclear Information System (INIS)

    Concrete cask spent fuel storage system is considered to essentially have an economical advantage and becoming widely used. For vertically free-standing concrete cask on the floor pad in the cask storage facility, its tipping-over and sliding behavior during earthquake is one of the technical key issues to guarantee its safe performance. In this paper, the experimental studies are reported by performing the excitation test with a scale model concrete cask using two-dimensional shaking table and the applicability of the energy spectrum approach is discussed. (author)

  17. Beneficial uses shipping system (BUSS) cask, safety analysis report for packaging: Volumes 1 and 2

    International Nuclear Information System (INIS)

    The Beneficial Uses Shipping System (BUSS) cask Safety Analysis Report for Packaging (SARP) was originally prepared by Sandia National Laboratory (SNL). After the certification process was completed, the ownership of the BUSS cask and associated SARP was transferred from SNL to the DOE Hanford site in Richland, Washington. During timely renewal of the BUSS cask certificate of compliance, the SARP was revised to (1) respond to the timely renewal questions, (2) consolidate the previous revision made by SNL, and (3) bring the SARP into compliance with the 1996 version of 10 CFR 71. Since the BUSS cask is now the responsibility of RL, the SARP was reissued as a Hanford document

  18. Evaluation of Equivalent Dose Rate of Interim Dry Storage Casks Loaded with Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Equivalent dose rate calculations of the CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks were performed using SCALE 4.3 computer codes system. These casks are planned for an interim storage of spent nuclear fuel at Ignalina NPP. The dose rate calculations were made on the sidelong, upper and lower surface of the cask and at the certain distance. Results show that dose rate values on the surface of the cask are much less then permissible value 1000 μSv/h when average burnup of fuel assembly is 20 GWd/tU. (author)

  19. Safety analysis report for packaging: the ORNL HFIR spent-fuel-element shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Eversole, R.E.; Just, R.A.; Llewellyn, G.H.

    1977-11-01

    The Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) spent-fuel-element shipping cask is used to transport HFIR, Oak Ridge Research Reactor (ORR), and other reactor fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures and tests were used to determine behavior of the cask relative to the general standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations.

  20. Bonner sphere neutron spectrometry at spent fuel casks

    CERN Document Server

    Rimpler, A

    2002-01-01

    For transport and interim storage of spent fuel elements from power reactors and vitrified highly active waste (HAW) from reprocessing, various types of casks are used. The radiation exposure of the personnel during transportation and storage of these casks is caused by mixed photon-neutron fields and, frequently, the neutron dose is predominant. In operational radiation protection, survey meters and even personal dosemeters with imperfect energy dependence of the dose-equivalent response are used, i.e. the fluence response of the devices does not match the fluence-to-dose equivalent conversion function. In order to achieve more accurate dosimetric information and to investigate the performance of dosemeters, spectrometric investigations of the neutron fields are necessary. Therefore, fluence spectra and dose rates were measured by means of a simple portable Bonner multisphere spectrometer (BSS). The paper describes briefly the experimental set-up and evaluation procedure. Measured spectra for different locat...

  1. Fuel Element Transfer Cask Modelling Using MCNP Technique

    Science.gov (United States)

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-01

    After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

  2. Stress analysis of closure bolts for shipping casks

    International Nuclear Information System (INIS)

    This report specifies the requirements and criteria for stress analysis of closure bolts for shipping casks containing nuclear spent fuels or high level radioactive materials. The specification is based on existing information conceming the structural behavior, analysis, and design of bolted joints. The approach taken was to extend the ASME Boiler and Pressure Vessel Code requirements and criteria for bolting analysis of nuclear piping and pressure vessels to include the appropriate design and load characteristics of the shipping cask. The characteristics considered are large, flat, closure lids with metal-to-metal contact within the bolted joint; significant temperature and impact loads; and possible prying and bending effects. Specific formulas and procedures developed apply to the bolt stress analysis of a circular, flat, bolted closure. The report also includes critical load cases and desirable design practices for the bolted closure, an in-depth review of the structural behavior of bolted joints, and a comprehensive bibliography of current information on bolted joints

  3. STACE: Source Term Analyses for Containment Evaluations of transport casks

    International Nuclear Information System (INIS)

    The development of the Source Term Analyses for Containment Evaluations (STACE) methodology provides a unique means for estimating the probability of cladding breach within transport casks, quantifying the amount of radioactive material released into the cask interior, and calculating the releasable radionuclide concentrations and corresponding maximum permissible leakage rates. Following the guidance of ANSI N14.5, the STACE methodology provides a technically defensible means for estimating maximum permissible leakage rates. These containment criteria attempt to reflect the true radiological hazard by performing a detailed examination of the spent fuel, CRUD, and residual contamination contributions to the releasable source term. The evaluation of the spent fuel contribution to the source team has been modeled fairly accurately using the STACE methodology. The structural model predicts the cask drop load history, the mechanical response of the fuel assembly, and the probability of cladding breach. These data are then used to predict the amount of fission gas, volitile species, and fuel fines that are releasable from the cask. There are some areas where data are sparse or lacking in which experimental validation is planned. Finally, the ANSI N14.5 recommendation that 3% and 100% of the fuel rods fail during normal and hypothetical accident conditions of transport, respectively, has been show to be overly conservative by several degrees of magnitude for these example analyses. Furthermore, the maximum permissible leakage rates for this example assembly under normal and hypothetical accident conditions are significanly higher that the leaktight requirements. By relaxing the maximum permissible leakage rates, the source term methodology is expected to significantly improvecask economics and safety

  4. Certification of a spent fuel cask for storage and transportation

    International Nuclear Information System (INIS)

    This paper addresses the US Nuclear Regulatory Commission's requirements for the dry storage and transportation of spent fuel, focusing on how the performance standards differ between storage and transportation. The paper also discusses the NRC cask review process, and some current issues in each area of certification. In addition, some of the issues associated with the US Department of Energy's proposed multi-purpose canister are discussed

  5. Structural evaluation of spent fuel dry storage cask

    International Nuclear Information System (INIS)

    In a various regulations and standards related to the spent fuel storage, the storage casks should be designed to sustain the structural integrity under the accident conditions of predicted operation and design criteria. These conditions for the structural evaluation requires the drop, tip-over, wind like tornado and typhoon, flood and earthquake. This paper describes the load cases and conceptual evaluation method for the structural evaluation. Preliminary safety analysis of the concrete storage system were performed

  6. Performance of bolted closure joint elastomers under cask aging conditions

    Energy Technology Data Exchange (ETDEWEB)

    Verst, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Skidmore, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Daugherty, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-23

    The bolted closure joint of a bare spent fuel cask is susceptible to age-related degradation and potential loss of confinement function under long-term storage conditions. Elastomeric seals, a component of the joint typically used to facilitate leak testing of the primary seal that includes the metallic seal and bolting, is susceptible to degradation over time by several mechanisms, principally via thermo-oxidation, stress-relaxation, and radiolytic degradation under time and temperature condition. Irradiation and thermal exposure testing and evaluation of an ethylene-propylene diene monomer (EPDM) elastomeric seal material similar to that used in the CASTOR® V/21 cask for a matrix of temperature and radiation exposure conditions relevant to the cask extended storage conditions, and development of semiempirical predictive models for loss of sealing force is in progress. A special insert was developed to allow Compressive Stress Relaxation (CSR) measurements before and after the irradiation and/or thermal exposure without unloading the elastomer. A condition of the loss of sealing force for the onset of leakage was suggested. The experimentation and modeling being performed could enable acquisition of extensive coupled aging data as well as an estimation of the timeframe when loss of sealing function under aging (temperature/radiation) conditions may occur.

  7. The Performance of Spent Fuel Casks in Severe Tunnel Fires

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission (NRC), working with the National Institute of Standards and Technology (NIST), Pacific Northwest National Laboratory (PNNL), and the National Transportation Safety Board (NTSB), performed analyses to predict the response of various spent fuel transportation cask designs when exposed to a fire similar to that which occurred in the Howard Street railroad tunnel in downtown Baltimore, Maryland on July 18, 2001. The thermal performance of three different spent fuel cask designs (HOLTEC HI-STAR 100, TransNuclear TN-68, and NAC-LWT) was evaluated with the ANSYSR and COBRA-SFS analysis codes, utilizing boundary conditions for the tunnel fire obtained using NIST's Fire Dynamics Simulator (FDS) code. NRC Staff evaluated the potential for a release of radioactive material from each of the three transportation casks analyzed for the Baltimore tunnel fire scenario. The results of these analyses are described in detail in Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario, NUREG/CR-6886, published in draft for comment in November 2005. Comments received by the NRC on NUREG/CR-6886 will be addressed in the final version of the report. (authors)

  8. Conceptual evaluation of metal storage cask for conditioned spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Seo, K. S.; Shin, H. S.; Lee, J. C.; Bang, K. S.; Kim, H. D.; Park, S. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2003-07-01

    The storage parameters of spent PWR fuel are radioactivity, heat power and its volume. Those values could be reduced to about a quarter by an Advanced spent fuel Conditioning Process (ACP). Firstly, a storage concept and scenario were established considering the characteristics of conditioned spent fuel. If the efficiency of the cooling system is improved and the appropriate quantities of the conditioned spent fuel are stored, the conditioned spent fuels could be stored at a four times higher level of spent fuel storage. One storage unit of conditioned spent fuel was designed to have its capacity equivalent to one PWR spent fuel. It was supposed that a metal storage cask has 7 baskets that can load 28 storage units. Those capacities means that 28 spent PWR fuels in metal casks can be stored. The conceptual evaluations of the critical, shielding, thermal and structural fields were performed. In conclusion, the results of the conceptual evaluations show that the proposed metal cask satisfied with the important design criteria at a smaller size than the existing systems.

  9. An analysis of contingencies for making casks available for use during the early years of Federal Waste Management System operations

    International Nuclear Information System (INIS)

    A study has been performed to examine the contingencies that could be pursued by the Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) for shipping spent fuel beginning in 1998. OCRWM's current plan is to initiate operations using early production units of Initiative I truck and rail/barge casks that are presently being designed. Contingencies to this plan were considered in case some unforeseen event occurs that precludes the Initiative I casks from entering into service early in 1998 in sufficient quantities (both numbers and types) to satisfy DOE's shipping needs. Specifically, the study addressed the potential availability of cask systems, selected several cask usage scenarios, determined the requirements for casks under these scenarios, generically assessed different strategies for acquiring casks or the use of casks, and generically assessed cask fabrication capabilities. Issues concerning both domestic and foreign resources were addressed with a focus on the first five years of Federal Waste Management System (FWMS) operation

  10. Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N. [Oak Ridge National Lab., Oak Ridge, TN (United States)

    2004-07-01

    Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO{sub 2}, Al{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO{sub 2}) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al{sub 2}O{sub 3}) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO{sub 2} for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate

  11. Al analysis and design of dry storage cask of spent nuclear fuel

    International Nuclear Information System (INIS)

    According to thermal analysis of the existing concrete cask, the maximum temperature occurred at the upper side of cask. If the cask lid is made of concrete, the temperature of concrete in lid exceeds the allowable value. Based on that result, research is progressed focusing on two strategies - one is to increase thermal margin, another is to decrease the lid concrete temperature. Here, thermally - enhanced design is suggested and analyzed. This design features the air flow duct in the lid and the thermal shielding disk installed between canister and lid. Air flow duct on the center of lid concrete connected to existing air outlet can decrease temperature by promoting the convection heat transfer, and thermal shielding disk bearing smaller hole located on the center can maintain the increased convection heat transfer and minimize radiation heat transfer from canister to lid concrete for the lid concrete temperature not to be over the allowable value. Thermal analysis result for this design shows that it can be very successful to achieve these objectives. The overall component of cask temperature decrease by 2-10 .deg. C, and the lid concrete temperature dropped from above 100 to 87.5 .deg. C which is below the allowable value 93 .deg. C. In addition, heat removal of cask depending on distance between casks was investigated. Cask heat is removed by convection and radiation heat transfer at an external surface to environment. Naturally, these heat transfers are mainly affected by ambient temperature. The ambient temperature can be affected if the thermal boundary layer is overlapped. So, thermal boundary layer thickness of cask was calculated to estimate to see if the ambient temperature is affected by other cask. Boundary layer thickness is calculated is too small just about 5cm. It is concluded that distance between casks can do little impact on heat removal of cask in a practical view

  12. Analisis SWOT Dalam Upaya Meningkatkan Keunggulan Bersaing Pada Salon Cleopatra

    OpenAIRE

    Syafitri, Yudhi

    2010-01-01

    Tujuan penelitian adalah untuk mengetahui dan menganalisis peranan analisis SWOT dalam upaya meningkatkan keunggulan bersaing pada Salon Cleopatra. Kesimpulan yang diperoleh penulis berdasarkan hasil dari penelitian yang dilakukan dengan menggunakan analisis SWOT dan Matriks SPACE dengan strategi agresif yang dilakukan oleh Salon Cleopatra sudah cukup baik, akan tetapi Salon Cleopatra melakukan strategi berdasarkan pengalaman, insting dan keadaan pasar, bukan dilakukan berdasarkan strategi a...

  13. 76 FR 33121 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-06-08

    ... 3150-AI90 List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition AGENCY: Nuclear... the NRC's spent fuel storage regulations to add the Holtec HI-STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage Casks'' as Certificate of Compliance Number 1032. DATES:...

  14. 76 FR 17037 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-03-28

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AI90 List of Approved Spent Fuel Storage Casks: HI.... Nuclear Regulatory Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations to add the HI-STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage...

  15. 78 FR 22411 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Amendment No. 8; Corrections

    Science.gov (United States)

    2013-04-16

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Amendment No... direct final rule amending its spent fuel storage regulations by revising the Holtec International, Inc. (Holtec) HI-STORM 100 Cask System listing within the ``List of Approved Spent Fuel Storage Casks''...

  16. Overview of research and development of metal cask for transport and storage of spent nuclear fuel in Japan

    International Nuclear Information System (INIS)

    The paper overviews experimental studies of dual-purpose metal casks carried out in Japan. Full-scale casks were dropped onto a reinforced concrete target simulating hypothetical accidental drop during handling procedure in a storage facility. In some cases, leakage from the primary lid was detected, but no leakage from the secondary lid. A heavy weight drop test was carried out onto a full-scale cask simulating hypothetical collapse of a storage building due to earthquake, etc. The cask maintained its integrity. A full-scale cask was covered with a thermal insulator simulating a hypothetical burial by debris due to a building collapse in earthquake, etc. Some components might need to be recovered from the debris before reaching their critical temperature. A scale-model of a cask was subjected to seismic motion on a shaking table simulating an earthquake. The cask was rocking more for an earthquake with longer wavelength. Long-term containment of metal gaskets in double lid structure of casks has been tested with full-scale lid model. Transportability of cask after long-term storage was tested simulating degradation of cask components. Effects of aging of cask body metal, basket metal, seal and neutron shielding materials were investigated. With those degradations, cask performance in terms of shielding, sub-criticality, heat removal and containment were investigated. (author)

  17. Sensitivity Analysis Applied to the Validation of the 10 B Capture Reaction in Nuclear Fuel Casks

    Energy Technology Data Exchange (ETDEWEB)

    Goluoglu, S

    2004-03-18

    Boron has commonly been used in nuclear fuel casks to ensure a sufficient margin of subcriticality. The amount of boron used in most casks far exceeds the amount of boron present in any of the available benchmark experiments. Such heavy loadings of boron in the casks may result in considerable spectral differences as compared to the benchmarks, resulting in boron sensitivities that are very different from those of the benchmarks. Before the calculations to determine the nuclear safety margin for various fuel loadings are deemed acceptable, as part of the safety basis, the computer code and cross sections must be validated against experimental benchmarks that cover the area of applicability of the proposed cask design. Therefore, this study was performed to determine if these available benchmarks can be used to validate a criticality code and neutron cross sections for the fuel casks. The sensitivity/uncertainty methodology has been applied to several application cask systems with different boron areal densities. Although, the sensitivities of the nuclear fuel cask applications are not completely covered by the set of benchmarks that were used in this study with regard to the 10B capture cross section, the effect of this lack of coverage on the keff is minimal. Thus, the experimental biases are determined to be appropriate for the cask systems, and no additional bias (penalty) due to high boron loading need be imposed.

  18. Modelling of RBMK-1500 SNF storage casks activation during very long term storage.

    Science.gov (United States)

    Narkunas, Ernestas; Smaizys, Arturas; Poskas, Povilas; Ragaisis, Valdas

    2016-09-01

    Existing interim spent nuclear fuel storage facility (SNFSF) at Ignalina nuclear power plant in Lithuania is fully loaded with CASTOR(®)RBMK-1500 and CONSTOR(®)RBMK-1500 storage casks. The planned lifetime of these casks is 50 years and the first loaded cask was moved to the SNFSF in 1999. The start of operation of disposal facility in Lithuania is foreseen later than the planned interim storage ends. So, the possibilities to extend the storage period over 50 years should be considered. Therefore, the casks decommissioning issues should be taken into account, as due to prolonged neutron irradiation casks materials could became activated. This paper presents modelling results of storage casks neutron activation during 300 year storage period. Modelling results show, that after 50 years of storage, side-wall and bottom of CASTOR(®)RBMK-1500 cask are activated above clearance criteria. However, for 100-300 year storage period all of the casks components could be free released. PMID:27344524

  19. 78 FR 16619 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-03-18

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AJ22 List of Approved Spent Fuel Storage Casks... Regulatory Commission (NRC) is proposing to amend its spent fuel storage regulations by revising the NAC... within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 3 to Certificate...

  20. 76 FR 70331 - List of Approved Spent Fuel Storage Casks: MAGNASTOR ® System, Revision 2

    Science.gov (United States)

    2011-11-14

    ... part 72, entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181... spent fuel storage cask designs. The NRC subsequently issued a final rule on November 21, 2008 (73 FR... COMMISSION 10 CFR Part 72 RIN 3150-AI91 List of Approved Spent Fuel Storage Casks: MAGNASTOR System,......

  1. 75 FR 42339 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

    Science.gov (United States)

    2010-07-21

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150--AI88 List of Approved Spent Fuel Storage Casks: NAC.... Nuclear Regulatory Commission (NRC) is proposing to amend its spent fuel storage cask regulations by... 72. PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL,...

  2. 75 FR 33678 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Science.gov (United States)

    2010-06-15

    ... COMMISSION 10 CFR Part 72 RIN 3150-AI86 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision... Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising the NAC International Inc. (NAC) MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks''...

  3. 77 FR 4203 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Science.gov (United States)

    2012-01-27

    ... 3150-AI91 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 2 AGENCY: Nuclear... amended the NRC's spent fuel storage regulations by revising the NAC International, Inc. (NAC) MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 2...

  4. Regulation of dopamine release by CASK-β modulates locomotor initiation in Drosophila melanogaster

    Directory of Open Access Journals (Sweden)

    Justin eSlawson

    2014-11-01

    Full Text Available CASK is an evolutionarily conserved scaffolding protein that has roles in many cell types. In Drosophila, loss of the entire CASK gene or just the CASK-β transcript causes a complex set of adult locomotor defects. In this study, we show that the motor initiation component of this phenotype is due to loss of CASK-β in dopaminergic neurons and can be specifically rescued by expression of CASK-β within this subset of neurons. Functional imaging demonstrates that mutation of CASK-β disrupts coupling of neuronal activity to vesicle fusion. Consistent with this, locomotor initiation can be rescued by artificially driving activity in dopaminergic neurons. The molecular mechanism underlying this role of CASK-β in dopaminergic neurons involves interaction with Hsc70-4, a molecular chaperone previously shown to regulate calcium-dependent vesicle fusion. These data suggest that there is a novel CASK-β-dependent regulatory complex in dopaminergic neurons that serves to link activity and neurotransmitter release.

  5. Licensing and safety issues associated with dry cask storage update. Panel Discussion

    International Nuclear Information System (INIS)

    Full text of publication follows: Panelists from the nuclear industry, cask vendors, the U.S. Department of Energy (DOE), and the U.S. Nuclear Regulatory Commission will speak to the current status of licensing casks for interim storage and shipping to the DOE permanent site and alternate interim private storage initiatives. Subject coverage will include a broad range of relevant issues. (authors)

  6. ANALISIS PERBANDINGAN SPAP, IAS DAN SPKN

    OpenAIRE

    Maylia Pramono Sari

    2012-01-01

    Dalam mewujudkan Good Goveronance  akan dilakukan analisis perbandingan mengenai standar pemeriksaan nasional yaitu SPAP, ISA, dan SPKN untuk menemukan standar yang lebih tepat dan lebih lengkap untuk pemeriksaan sektor publik. Persamaan SPKN dan ISS  adalah tanggungjawab manajemen dan tanggungjawab auditor sedangkan perbedaannya adalah badan yang menerbitkannya. Perbedaan yang mendasar terletak pada karakter pemeriksaan BPK. Karakter tersebut adalah keharusan pemeriksa BPK untuk merancang ...

  7. Seismic Response Analysis of Spent Nuclear Fuel Metal Storage Cask considering Soil- Structure Interaction Effects

    Energy Technology Data Exchange (ETDEWEB)

    Baeg, Chang-Yeal; Lee, Kyung-Ho; Lee, Dae-Ki [Nuclear Engineering and Technology Institute, Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Jung, In-Su; Song, Won-Tae; Jin, Han-Uk; Kim, Jong-Soo [KONES, Seoul (Korea, Republic of)

    2008-05-15

    Maintaining of the structure safety for the metal storage cask is important to store spent nuclear fuel under a seismic events. Sliding and overturning behavior must be estimated because the metal cask systems are to be installed as free standing structures on reinforced concrete pads. This behavior can cause a serious problem in the integrity of spent nuclear fuel by the impact between neighboring casks. Also, soil condition should be considered since the cask's behavior is strongly affected by the characteristics of the base soil condition. In this study, the seismic response analysis was carried out in order to evaluate the behavior of the metal storage cask under earthquake envelopment considering Soil-Structure Interaction (SSI) effects.

  8. A robotic system to conduct radiation and contamination surveys on nuclear waste transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Harrigan, R.W.; Sanders, T.L.

    1990-06-01

    The feasibility of performing, numerous spent fuel cask operations using fully integrated robotic systems is under evaluation. Using existing technology, operational and descriptive software and hardware in the form of robotic end effectors are being designed in conjunction with interfacing cask components. A robotic radiation and contamination survey system has been developed and used on mock-up cask hardware to evaluate the impact of such fully automated operations on cask design features and productivity. Based on experience gained from the survey system, numerous health physics operations can be reliably performed with little human intervention using a fully automated system. Such operations can also significantly reduce time requirements for cask-receiving operations. 7 refs., 51 figs., 6 tabs.

  9. Effectively meeting spent fuel storage needs with a family of dry storage casks

    International Nuclear Information System (INIS)

    During 1988--89, a number of nuclear utilities have announced their intent of developing supplemental spent fuel storage. These on-site facilities are to be operable by 1991--93. This paper discusses how the Castor ductile cast iron (DCI) storage casks is a tested and licensed means of meeting this fuel storage need. Since 1986, a total of 14 casks have been sold to the Virginia Power Co. (V.P.). Eight casks are now loaded and in storage at the V.P. Surry Nuclear Station. These casks are directly pool loaded and moved to a storage pad using straight forward handling operations. Once on the pad, there is no further need for cask operation or maintenance with this sealed and passive storage system

  10. The dry storage cask in interim storage facility and safeguards activity

    International Nuclear Information System (INIS)

    The Japan Atomic Power Company (JAPC) is preparing for interim storage of spent fuel at Recyclable-Fuel Storage Center (RFSC) in Aomori Prefecture. Metallic dry casks are employed to contain the spent fuel and to serve for about 50 years in RFSC. Metallic dry casks have already been used for spent fuel dry storage at Tokai No.2 power station. But, RFSC is not exactly the same as the dry storage facility in Tokai No.2 power station, for example, casks are transported out side of the reactor site and RFSC has no fuel handling system. Therefore, additional implementation of safeguards is necessary. This report introduces the design and handling of metallic dry casks for RFSC and the currently developing status of the safeguards activity such as containment and surveillance for the fuel loading at the power station, the cask receipt and storage at RFSC, etc. (author)

  11. A robotic system to conduct radiation and contamination surveys on nuclear waste transport casks

    International Nuclear Information System (INIS)

    The feasibility of performing, numerous spent fuel cask operations using fully integrated robotic systems is under evaluation. Using existing technology, operational and descriptive software and hardware in the form of robotic end effectors are being designed in conjunction with interfacing cask components. A robotic radiation and contamination survey system has been developed and used on mock-up cask hardware to evaluate the impact of such fully automated operations on cask design features and productivity. Based on experience gained from the survey system, numerous health physics operations can be reliably performed with little human intervention using a fully automated system. Such operations can also significantly reduce time requirements for cask-receiving operations. 7 refs., 51 figs., 6 tabs

  12. CLASSIFICATION OF THE MGR CARRIER/CASK HANDLING SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    J.A. Ziegler

    2001-02-08

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) carried cask handling system structures, systems and components (SSCs) performed by the MGR Preclosure Safety and Systems Engineering Section. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 2000). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 2000).

  13. Safety analysis report for packaging (onsite) multicanister overpack cask

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.S.

    1997-07-14

    This safety analysis report for packaging (SARP) documents the safety of shipments of irradiated fuel elements in the MUlticanister Overpack (MCO) and MCO Cask for a highway route controlled quantity, Type B fissile package. This SARP evaluates the package during transfers of (1) water-filled MCOs from the K Basins to the Cold Vacuum Drying Facility (CVDF) and (2) sealed and cold vacuum dried MCOs from the CVDF in the 100 K Area to the Canister Storage Building in the 200 East Area.

  14. Safety analysis report for packaging (onsite) Castor GSF cask

    International Nuclear Information System (INIS)

    The CASTOR GSF packaging was designed and fabricated to be a certified Type B(U) packaging and comply with the requirements of the International Atomic Energy Agency (IAEA) for transport of up to five sealed canisters of vitrified radioactive materials. This onsite Safety Analysis Report for Packaging (SARP) provides the analysis and evaluations necessary to demonstrate that the casks, with the canister payload, meet the intent of the Type B packaging regulations set forth in 10 CFR 71 and therefore meet the onsite transportation safety requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping

  15. Safety analysis report for packaging (onsite) multicanister overpack cask

    International Nuclear Information System (INIS)

    This safety analysis report for packaging (SARP) documents the safety of shipments of irradiated fuel elements in the MUlticanister Overpack (MCO) and MCO Cask for a highway route controlled quantity, Type B fissile package. This SARP evaluates the package during transfers of (1) water-filled MCOs from the K Basins to the Cold Vacuum Drying Facility (CVDF) and (2) sealed and cold vacuum dried MCOs from the CVDF in the 100 K Area to the Canister Storage Building in the 200 East Area

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

    International Nuclear Information System (INIS)

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

  17. Analysis and design of dry cask storage pads for plant hatch Isfsi

    International Nuclear Information System (INIS)

    An independent spent fuel storage installation (ISFSI) at Southern Company's Edwin I. Hatch Nuclear Plant (HNP) was completed, licensed, and put in service in the summer of 2000. Currently this dry cask on-site storage facility provides a temporary spent fuel storage for three Holtec HI-STAR 100 system casks. After re-racking and rod consolidation efforts, the HNP ISFSI was necessary to maintain a full core discharge capacity of its spent nuclear fuel pools and also to temporarily delay a need for a permanent off-site spent nuclear fuel repository. The HNP ISFSI was carried out to meet the following three main criteria established at the beginning of the HNP Spent Fuel Storage Project. These three criteria were 1) to use the general license approach which utilizes the license of the cask vendor rather than obtaining a site-specific license, 2) to select only dry cask products that are intended for dual purpose licensing, and 3) to acquire sufficient dry cask storage capacity to fully meet the plant's need. This paper describes the major steps of analysis and design of dry cask storage pads for Plant Hatch ISFSI. Results showed that HNP ISFSI met the applicable codes, regulatory and cask vendor requirements. (author)

  18. Documentation for initial testing and inspections of Beneficial Uses Shipping System (BUSS) Cask

    International Nuclear Information System (INIS)

    The purpose of this report is to compile data generated during the initial tests and inspections of the Beneficial Uses Shipping System (BUSS) Cask. In addition, this report will verify that the testing criteria identified in section 8.1 of the BUSS Cask Safety Analysis Report for Packaging (SARP) was met. The BUSS Cask Model R-1 is a type B shipping container used for shipment of radioactive cesium-137 and strontium-90 capsules to Waste Encapsulation and Storage Facility (WESF). The BUSS Cask body and lid are each one-piece forgings fabricated from ASTM A473, Type 304 stainless steel. The primary purpose of the BUSS Cask is to provide shielding and confinement as well as impact, puncture, and thermal protection for the capsules under both normal and accident conditions. Chapter 8 of the BUSS Cask SARP requires several acceptance tests and inspections, each intended to evaluate the performance of different components of the BUSS Cask system, to be performed before its first use. The results of the tests and inspections required are included in this document

  19. Implications of the Baltimore Rail Tunnel Fire for Full-Scale Testing of Shipping Casks

    Energy Technology Data Exchange (ETDEWEB)

    Halstead, R. J.; Dilger, F.

    2003-02-25

    The U.S. Nuclear Regulatory Commission (NRC) does not currently require full-scale physical testing of shipping casks as part of its certification process. Stakeholders have long urged NRC to require full-scale testing as part of certification. NRC is currently preparing a full-scale casktesting proposal as part of the Package Performance Study (PPS) that grew out of the NRC reexamination of the Modal Study. The State of Nevada and Clark County remain committed to the position that demonstration testing would not be an acceptable substitute for a combination of full-scale testing, scale-model tests, and computer simulation of each new cask design prior to certification. Based on previous analyses of cask testing issues, and on preliminary findings regarding the July 2001 Baltimore rail tunnel fire, the authors recommend that NRC prioritize extra-regulatory thermal testing of a large rail cask and the GA-4 truck cask under the PPS. The specific fire conditions and other aspects of the full-scale extra-regulatory tests recommended for the PPS are yet to be determined. NRC, in consultation with stakeholders, must consider past real-world accidents and computer simulations to establish temperature failure thresholds for cask containment and fuel cladding. The cost of extra-regulatory thermal testing is yet to be determined. The minimum cost for regulatory thermal testing of a legal-weight truck cask would likely be $3.3-3.8 million.

  20. Two decades of experience with more than 750 CASTOR and CONSTOR transport and storage casks

    International Nuclear Information System (INIS)

    In 1983 the world-wide first dual purpose transport and storage cask - a CASTOR registered Ic-DIORIT - was loaded in Wuerenlingen/ Switzerland. Meanwhile CASTOR registered casks are used at 24 sites on four continents. Spent fuel assemblies of PWR, BWR, VVER, RBMK, FBR, MTR and THTR as well as vitrified high active waste canisters are transported and/or stored in these kinds of monolithic metal casks. MOX spent fuel of PWR and BWR has been loaded, too. Starting in the mid of the 90s, GNB developed the new CONSTOR registered cask concept, which is based on a double liner technology with a layer of heavy concrete as shielding material inbetween. This CONSTOR registered cask concept fulfils all design criteria for transport and for storage given by the IAEA recommendations and by national authorities. Up to now, more than 750 CASTOR registered and CONSTOR registered casks have been used for transports or/and loaded for longterm interim storage. More than two decades of storage experience attest to the excellent behavior of the casks including the metallic gaskets and the tightness monitoring system. Detailed measurements of temperatures and of gamma and neutron dose rates have shown in each case that the safety requirements have been fulfilled. These measurements allowed to reduce unnecessary safety margins to optimize the benefit for the user

  1. Contamination transfers during fuel transport cask loading. A concrete situation

    Energy Technology Data Exchange (ETDEWEB)

    Fournel, B.; Turchet, J.P.; Faure, S.; Allinei, P.G. [DEN/DED Centre d' Etudes de Cadarache, 13 - Saint Paul lez Durance (France); Briquet, L. [EDF GENV, 93 - Saint Denis (France); Baubet, D. [SGS Qualitest Industrie, 30 - Pont Saint Esprit (France)

    2002-07-01

    In 1998, a number of contamination cases detected during fuel shipments have been pointed out by the french nuclear safety authority. Wagon and casks external surfaces were partly contaminated upon arrival in Valognes railway terminal. Since then, measures taken by nuclear power plants operators in France and abroad solved the problem. In Germany, a report analyzing the situation in depth has been published in which correctives actions have been listed. In France, EDF launched a large cleanliness program (projet proprete radiologique) in order to better understand contamination transfers mechanisms during power plants exploitation and to list remediation actions to avoid further problems. In this context, CEA Department for Wastes Studies at Cadarache (CEA/DEN/DED) was in charge of a study about contamination transfers during fuel elements loading operations. It was decided to lead experiments for a concrete case. The loading of a transport cask at Tricastin-PWR-1 was followed in november 2000 and different analysis comprising water analysis and smear tests analysis were carried out and are detailed in this paper. Results are discussed and qualitatively compared to those obtained in Philippsburg-BWR, Germany for a similar set of tests. (authors)

  2. Pilot study dismantlement of 20 lead-lined shipping casks

    International Nuclear Information System (INIS)

    This report describes a pilot study conducted at the INEL to dismantle lead-lined casks and shielding devices, separate the radiologically contaminated and hazardous materials, and recycle resultant scrap lead. The facility areas where the work was performed, dismantlement methods, and process equipment are described. Issues and results associated with recycling the lead as a free-released scrap metal are presented and discussed. Data and results from the pilot study are summarized and presented. The study concluded that cask dismantlement at the INEL can be performed as a legitimate recycling activity for scrap lead. Ninety-one percent of the lead recovered passed free-release criteria. The value of the 50,375 lb of recovered lead is approximately $0.45/lb. Resultant waste streams can be satisfactorily treated and disposed. Only very low levels of bulk radiological contamination (47 picocuries/gram of 137 Cs and 3.2 picocuries/gram of 6OCo) were detected in the lead rejected for free release

  3. Transfer cask system design activities: status and plan

    Energy Technology Data Exchange (ETDEWEB)

    Locke, D., E-mail: darren.locke@f4e.europa.eu [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Gutierrez, C. Gonzalez; Damiani, C.; Gracia, V. [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Friconneau, J.-P.; Martins, J.-P.; Blight, J. [ITER Organisation, CS 90 046, 13067St. Paul Lez Durance Cedex (France)

    2011-10-15

    The ITER Cask and Plug Remote Handling System (CPRHS), a.k.a. Transfer Cask System, is a critical element of the ITER Remote Maintenance System (IRMS) devoted to transportation of components between the Tokamak building and Hot Cell. Due to the necessary confinement of contaminated components the CPRHS is defined as Safety Importance Class 1 (SIC-1) plus the mobile nature of the CPRHS brings with it a significant number of complex interfaces with other ITER sub-systems. With a total CPRHS fleet in excess of 20 units, including seven typologies, the management of design and procurement needs to be carefully planned and implemented to ensure compliance with ITER's requirements. Fusion for Energy (F4E) and its beneficiaries/contractors are currently working under ITER Task Agreements (ITAs) on the conceptual design of the CPRHS and, following the signing of the Procurement Arrangement (PA) in mid 2012, will take responsibility for the entire CPRHS fleet. F4E must, therefore, develop a robust strategy to meet the needs of both ITER machine assembly (for which a number of CPRHS units will be utilised) and the remote maintenance of ITER. Within this context this paper will present the status of the current CPRHS design activities, highlight some of the significant issues which will be faced during procurement and present the overall strategy which is being implemented by F4E in order to meet these challenging objectives.

  4. Spent nuclear fuel shipping cask handling capabilities of commercial light water reactors

    International Nuclear Information System (INIS)

    This report describes an evaluation of the cask handling capabilities of those reactors which are operating or under construction. A computerized data base that includes cask handling information was developed with information from the literature and utility-supplied data. The capability of each plant to receive and handle existing spent fuel shipping casks was then evaluated. Modal fractions were then calculated based on the results of these evaluations and the quantities of spent fuel projected to be generated by commercial nuclear power plants through 1998. The results indicated that all plants are capable of receiving and handling truck casks. Up to 118 out of 130 reactors (91%) could potentially handle the larger and heavier rail casks if the maximum capability of each facility is utilized. Design and analysis efforts and physical modifications to some plants would be needed to achieve this high rail percentage. These modifications would be needed to satisfy regulatory requirements, increase lifting capabilities, develop rail access, or improve other deficiencies. The remaining 12 reactors were determined to be capable of handling only the smaller truck casks. The percentage of plants that could receive and handle rail casks in the near-term would be reduced to 64%. The primary reason for a plant to be judged incapable of handling rail casks in the near-term was a lack of rail access. The remaining 36% of the plants would be limited to truck shipments. The modal fraction calculations indicated that up to 93% of the spent fuel accumulated by 1998 could be received at federal storage or disposal facilities via rail (based on each plant's maximum capabilities). If the near-term cask handling capabilities are considered, the rail percentage is reduced to 62%

  5. Drop accident analyses of dry metal cask without impact limiter and evaluation of leak rate

    International Nuclear Information System (INIS)

    In Japan, utilities are preparing to commence a private Interim Storage Facility (ISF) of spent fuels in Mutucity, Aomori prefecture from 2010. In the ISF, metal casks for transport and storage will be adopted and handled without an impact limiter. Cask drop tests without the impact limiter using an actual size simulated cask were carried out by CRIEPI (Central Research Institute of Electric Power Industry) in 2005. Then cases of cask drop tests were analyzed and the leak rate characteristics of a metal gasket were investigated. A general non-linear dynamic simulation computer code LS-DYNA was used in analyses. The collision velocity of the cask was calculated assuming free drop from an initial position for both horizontal drop and rotational drop. Although the drop height was 1 m in the tests, it was changed to 1.5 m and 2.0 m as parameters in the calculation for investigation of the leak rate characteristic. It was supposed that the increase of the leak rate was not only due to an increase of the total sliding movement of the lid but also caused by plastic deformation of flange or bolts. A correlation curve between total sliding movement of lid and leak rate was settled for leak rate of cask drops without the impact limier, based on results of the previous test using small-scale sized model (small scale test). Under these postulations, the leak rate could be evaluated by the correlation curve and obtained total sliding movement of the lid. In the simulated cask used for the test, a clearance between the lid and the cask body was small and the total sliding movement was limited. The leak rate estimation methodology would be applicable to the actual cask drop accident without the impact limiter, if the plastic deformation were not occurred at the flange. (author)

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

    International Nuclear Information System (INIS)

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

  7. Material specification and quality control program for ductile iron spent fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Rehmer, B.; Frenz, H.; Weidlich, S.; Kuehn, H.D.

    1995-12-31

    In the process of testing spent fuel casks, BAM is gaining a lot of relevant data regarding the quality level of Ductile Cast Iron (DCI). This paper discusses the basic parameters governing the material behavior of ferritic and ferritic-pearlitic DCI and reviews the development of cask quality over the last years. The effect of microstructure and sample size on the fracture toughness of DCI is discussed. The results of a test program show the prominent effect of pearlite content and graphite nodule structure in the mechanical and fracture toughness characteristics of DCI. This observation is important for quality assurance programs for shipping and storage casks of radioactive materials.

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

    International Nuclear Information System (INIS)

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

  9. CASKET: a computer code system for thermal and structural analyses of radioactive material transport and/or storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-05-01

    A computer code system CASKET (CASK thermal and structural analyses and Evaluation code system) for the thermal and structural analyses which are indispensable for radioactive material transport and/or storage cask designs has been developed. The CASKET is a simplified computer code system to perform parametric analyses on sensitivity evaluations in designing a cask and conducting its safety analysis. Main features of the CASKET are as follow: (1) it is capable to perform impact analysis of casks with shock absorbers, (2) it is capable to perform impact analysis of casks with fins. (3) puncture analysis of casks is capable, (4) rocking analysis of casks during seismic load is capable, (5) material property data library are provided for impact analysis of casks, (6) material property data library are provided for thermal analysis of casks, (7) fin energy absorption data library are provided for impact analysis of casks with fins are and (8) not only main frame computers (OS MSP) but also work stations (OS UNIX) and personal computers (OS Windows 3.1) are available. In the paper, brief illustrations of calculation methods are presented. Some calculation results are compared with experimental ones to confirm the computer programs are useful for thermal and structural analyses. (author)

  10. CASK stabilizes neurexin and links it to liprin-α in a neuronal activity-dependent manner.

    Science.gov (United States)

    LaConte, Leslie E W; Chavan, Vrushali; Liang, Chen; Willis, Jeffery; Schönhense, Eva-Maria; Schoch, Susanne; Mukherjee, Konark

    2016-09-01

    CASK, a MAGUK family protein, is an essential protein present in the presynaptic compartment. CASK's cellular role is unknown, but it interacts with multiple proteins important for synapse formation and function, including neurexin, liprin-α, and Mint1. CASK phosphorylates neurexin in a divalent ion-sensitive manner, although the functional relevance of this activity is unclear. Here we find that liprin-α and Mint1 compete for direct binding to CASK, but neurexin1β eliminates this competition, and all four proteins form a complex. We describe a novel mode of interaction between liprin-α and CASK when CASK is bound to neurexin1β. We show that CASK phosphorylates neurexin, modulating the interaction of liprin-α with the CASK-neurexin1β-Mint1 complex. Thus, CASK creates a regulatory and structural link between the presynaptic adhesion molecule neurexin and active zone organizer, liprin-α. In neuronal culture, CASK appears to regulate the stability of neurexin by linking it with this multi-protein presynaptic active zone complex. PMID:27015872

  11. TRANSPORTATION CASK RECEIPT/RETURN FACILITY CRITICALITY SAFETY EVALUATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-26

    The purpose of this design calculation is to demonstrate that the handling operations of transportation casks performed in the Transportation Cask Receipt and Return Facility (TCRRF) and Buffer Area meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC [Bechtel SAIC Company] 2004 [DIRS 171599], Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''Transportation Cask Receipt/Return Facility Description Document'' (BSC 2004 [DIRS 170217], Section 3.2.3). Specific scope of work contained in this activity consists of the following items: (1) Evaluate criticality effects for both dry and fully flooded conditions pertaining to TCRRF and Buffer Area operations for defense in depth. (2) Evaluate Category 1 and 2 event sequences for the TCRRF as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). This evaluation includes credible fuel reconfiguration conditions. In addition to the scope of work listed above, an evaluation was also performed of modeling assumptions for commercial spent nuclear fuel (CSNF) regarding inclusion of plenum and end regions of the active fuel. This calculation is limited to CSNF and US Department of Energy (DOE) SNF. it should be mentioned that the latter waste form is evaluated more in depth in the ''Canister Handling Facility Criticality Safety Calculations (BSC 2004 [DIRS 167614]). Further, the design and safety analyses of the naval SNF canisters are the responsibility of the US Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the TCRRF and Buffer Area and may not reflect the ongoing design evolution of the facility. However, it is anticipated that design changes to the facility layout will

  12. Desain Dan Analisis Perhitungan Roda Pendaratan Pesawat Tanpa Awak

    OpenAIRE

    Aulia, T. Muhammad Rinaldi

    2015-01-01

    Landing gear merupakan struktur pesawat yang berfungsi menahan beban statis pesawat dan juga beban dinamis ketika pesawat melakukan pendaratan. Dalam mendesain landing gear dilakukan pemilihan jenis landing gear dan dilakukan analisis perhitungan pada tiap komponen landing gear yang meliputi pusat gravitasi, tinggi pesawat, wheel base, wheel track, dan roda. Desain dan analisis perhitungan dilakukan dengan metode studi pustaka dimana setiap desain dan perhitungan didasarkan pada literatur pu...

  13. Nasionalisme dalam Novel 5 cm. Karya Donny Dhirgantoro: Analisis Strukturalisme

    OpenAIRE

    Ichwan, Yasir

    2014-01-01

    Penelitian ini menganalisis tentang struktur yang membangun nilai nasionalisme dan bentuk nasionalisme dalam novel 5 cm. karya Donny Dhirgantoro. Adapun tujuan penelitian ini adalah untuk memahami makna nasionalisme secara lebih luas. Metode pengumpulan data yang dilakukan peneliti adalah teknik pustaka, baca, dan catat. Teknik analisis data yang dilakukan peneliti adalah menganalisis unsur-unsur pembangun novel, mengaitkan antara unsur pembangun novel, menyajikan hasil analisis, dan menyimpu...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-11-01

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

  15. Neutronics and dose calculation for prospective spent nuclear fuel cask for Ghana Research Reactor - 1 facility

    International Nuclear Information System (INIS)

    Ghana Research Reactor-1 core is to be converted from highly enrich Uranium (HEU) fuel to low enriched Uranium (LEU) fuel in the near future: a storage cask will be needed to store the HEU fuel. Notwithstanding the core conversion process, It is also important for the facilitv to have a storage cask ready when the fuel is finally spent to temporarily store the fuel until permanent storage is provided. Winfrith Improved Multigroup Scheme-Argonne National Laboratory (WIMS-ANL). Reactor Burnup System (REBUS). Oak Ridge Isotope Generation (ORIGEN2) and Monte Carlo ''N'' Particle (MCNP5) codes have been used to design the cask. WIMS-ANL was used in generating cross sections for the REBUS code which was used in the burnup calculations. The REBUS code was used to estimate the core life time. An estimated core life of approximatcly 750 full-power-equivaicnt-days was obtained for reactor operation of 2hours a day. 4 days a week and 48 weeks in a year. The ORIGIN2 code recorded U-235 burnup weight percent of 2.90% whilst the result from the REBUS3 code was 2.86%. The amount of Pu-239 at the end of the irradiation period was 145 mg which is very low relative to other low power reactors. Isotopic inventory obtained from the ORIGIN2 and REBUS3 runs were used in setting up the MCNP5 input deck for the MCNP5 calculation of the criticality and dose rate. Six cask design options were investigated. The materials for the casks designs were selected based on their attenuation coefficient properties and their high removal cross section properties. The various materials were arranged in no specific order in multilayered casks. The reason for investigating six casks was to look at various arrangements of the cask layers that will optimize effective shielding. The spent nuclear fuel at discharge was used as the radioactivity source during the MCNP simulation. The multilayer cask shield comprise of serpentine concrete of density 5.14 g/cm3 and thickness 21.94cm which

  16. Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

    International Nuclear Information System (INIS)

    The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system

  17. Long term containment performance test for spent fuel transport/storage casks

    International Nuclear Information System (INIS)

    The use of transport/storage cask for spent fuel storage is considered to be rational and economical. Since the storage duration may continue for 40 years or so, the function of sealing radioactive materials in the casks must be reliable for long-term. Long-term containment test of full-scale spent fuel transport/storage cask models have been in progress since 1990 in CRIEPI, Japan. It has been 11 years since it started. The results so far demonstrate and confirm very reliable containment performance of the cask lid structure with metal gaskets. Using the test data it is predicted by Larson-Miller Parameter (LMP) method that the containment system will keep its integrity at least for 40 years. (author)

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

    International Nuclear Information System (INIS)

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

  19. Regulators Experiences in Licensing and Inspection of Dry Cask Storage Facilities

    International Nuclear Information System (INIS)

    The United States Nuclear Regulatory Commission (NRC), through the combination of a rigorous licensing and inspection program, ensures the safety and security of dry cask storage. NRC authorizes the storage of spent fuel at an independent spent fuel storage installation (ISFSI) under two licensing options: site-specific licensing and general licensing. In July 1986, the NRC issued the first site-specific license to the Surry Nuclear Power Plant in Virginia authorizing the interim storage of spent fuel in a dry storage cask configuration. Today, there are over 30 ISFSIs currently licensed by the NRC with over 700 loaded dry casks. Current projections identify over 50 ISFSIs by the year 2010. No releases of spent fuel dry storage cask contents or other significant safety problems from the storage systems in use today have been reported. This paper discusses the NRC licensing and inspection experiences. (authors)

  20. Storage and transportation of spent fuel and high-level waste using dry storage casks

    International Nuclear Information System (INIS)

    This paper describes the REA 2023 dry storage cask which has been designed for on-site storage and transportation of spent fuel and high-level waste. The REA 2023 is the first domestic commercial spent fuel dry storage cask completed for the Department of Energy program for demonstration of methods to improve on site utility fuel storage capacity. A description of the operations required for on-site handling and storage is provided with illustrations and photographs of the fabricated cask. An auxiliary skid is also described which is designed for both on-site handling/storage and transportation. A description of the lifting yoke and transportation impact limiters completes the total system for storage and transportation of spent fuel and high level waste in the REA 2023 casks

  1. Regulatory body experiences in licensing and inspection of dry cask storage facilities

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission (NRC), through a rigorous licensing and inspection programme, ensures the safety and security of dry cask storage. The NRC authorizes the storage of spent fuel at an independent spent fuel storage installation (ISFSI) under two licensing options: site specific licensing and general licensing. In July 1986, the NRC issued the first site specific licence to the Surry Nuclear Power Plant in Virginia, authorizing the interim storage of spent fuel in a dry storage cask configuration. Presently, there are over 40 ISFSIs licensed by the NRC, with over 800 loaded dry casks. Current projections indicate that there will be over 50 ISFSIs by the year 2010. No releases of spent fuel dry storage cask contents or other significant safety problems from the storage systems in use today have been reported. The paper discusses the NRC's licensing and inspection experiences. (author)

  2. Drop accident analyses of dry metal cask without impact limiter and evaluation of leak rate

    International Nuclear Information System (INIS)

    In Japan, utilities are preparing to initiate an independent interim storage facility (ISF) for spent fuel at Mutusi-city in Aomori prefecture in 2010. In the ISF, dual purpose metal casks which are used for both transportation and storage will be adopted, because no direct handling of spent fuel is necessary at the ISF, thereby reducing risks. The metal cask will be handled without impact limiters in the ISF. Therefore, supposing a hypothesis cask drop accident without the limiter, cask drop tests using an actual size simulated cask were analyzed and the leak characteristics from the flange with the metal gasket were investigated. The tests were conducted without the limiter, and the conditions were a horizontal drop and rotational impact with the supporting point at a trunnion. Before the calculation of this cask drop event, based on examination of results obtained from small scale tests for seal performance of flange with aged metal gasket, a correlation curve between total sliding movement of lid and leak rate was obtained. The relation between the total sliding movement of the lid and the leak rate obtained from the cask dropping tests without the impact limiter was compared with the correlation. Considering the leak rate increase due to aging of the gasket which is assumed to be ranging from 100 to 1000, the result from the cask drop tests agreed to the correlation with a 95% confidence level. Then, a general non-linear dynamic simulation computer code, LS-DYNA was used in the calculation of the cask drop tests. In the calculation, a half of the cask, considering axial symmetry, and a concrete floor were modelled. The calculation for the horizontal drop test was initiated just before a trunnion impacts the floor. For the rotational impact test, the calculation was initiated just before the edge of the outer flange impacting the floor. The impacting velocity of the cask was calculated assuming a free drop from the original position for both horizontal drop and

  3. Regulators experiences in licensing and inspection of dry cask storage facilities

    International Nuclear Information System (INIS)

    The United States Nuclear Regulatory Commission (NRC), through the combination of a rigorous licensing and inspection program, ensures the safety and security of dry cask storage. NRC authorizes the storage of spent fuel at an independent spent fuel storage installation (ISFSI) under two licensing options: site-specific licensing and general licensing. In July 1986, the NRC issued the first site specific license to the Surry Nuclear Power Plant in Virginia authorizing the interim storage of spent fuel in a dry storage cask configuration. Today, there are over 30 ISFSIs currently licensed by the NRC with over 700 loaded dry casks. Current projections identify over 50 ISFSIs by the year 2010. No releases of spent fuel dry storage cask contents or other significant safety problems from the storage systems in use today have been reported. This paper discusses the NRC licensing and inspection experiences. (author)

  4. Spent Fuel Transportation Cask Response to the Caldecott Tunnel Fire Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Adkins, Harold E.; Koeppel, Brian J.; Cuta, Judith M.

    2007-01-01

    On April 7, 1982, a tank truck and trailer carrying 8,800 gallons of gasoline was involved in an accident in the Caldecott tunnel on State Route 24 near Oakland, California. The tank trailer overturned and subsequently caught fire. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook analyses to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by truck. The Fire Dynamics Simulator (FDS) code developed by National Institute of Standards and Technology (NIST) was used to determine the thermal environment in the Caldecott tunnel during the fire. The FDS results were used to define boundary conditions for a thermal transient model of a truck transport cask containing spent nuclear fuel. The Nuclear Assurance Corporation (NAC) Legal Weight Truck (LWT) transportation cask was selected for this evaluation, as it represents a typical truck (over-the-road) cask, and can be used to transport a wide variety of spent nuclear fuels. Detailed analysis of the cask response to the fire was performed using the ANSYS® computer code to evaluate the thermal performance of the cask design in this fire scenario. This report describes the methods and approach used to assess the thermal response of the selected cask design to the conditions predicted in the Caldecott tunnel fire. The results of the analysis are presented in detail, with an evaluation of the cask response to the fire. The staff concluded that some components of smaller transportation casks resembling the NAC LWT, despite placement within an ISO container, could degrade significantly. Small transportation casks similar to the NAC LWT would probably experience failure of seals in this severe accident scenario. USNRC staff evaluated the radiological consequences of the cask response to the Caldecott tunnel fire. Although some

  5. Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    None

    1990-02-01

    This Preliminary Design Report (PDR) provides a detailed description of the design, analyses, and testing programs for the BR-100 cask. The BR-100 is a Type B(U) cask designed for transport by rail or barge. This report presents the preliminary analyses and tests which have been performed for the BR-100 and outlines the confirmatory analyses and tests which will be performed.

  6. DESIGN OF A CONCRETE SLAB FOR STORAGE OF SNF AND HLW CASKS

    Energy Technology Data Exchange (ETDEWEB)

    J. Bisset

    2005-02-14

    This calculation documents the design of the Spent Nuclear Fuel (SNF) and High-Level Waste (HLW) Cask storage slab for the Aging Area. The design is based on the weights of casks that may be stored on the slab, the weights of vehicles that may be used to move the casks, and the layout shown on the sketch for a 1000 Metric Ton of Heavy Metal (MTHM) storage pad on Attachment 2, Sht.1 of the calculation 170-C0C-C000-00100-000-00A (BSC 2004a). The analytical model used herein is based on the storage area for 8 vertical casks. To simplify the model, the storage area of the horizontal concrete modules and their related shield walls is not included. The heavy weights of the vertical storage casks and the tensile forces due to pullout at the anchorages will produce design moments and shear forces that will envelope those that would occur in the storage area of the horizontal modules. The design loadings will also include snow and live loads. In addition, the design will also reflect pertinent geotechnical data. This calculation will document the preliminary thickness and general reinforcing steel requirements for the slab. This calculation also documents the initial design of the cask anchorage. Other slab details are not developed in this calculation. They will be developed during the final design process. The calculation also does not include the evaluation of the effects of cask drop loads. These will be evaluated in this or another calculation when the exact cask geometry is known.

  7. Evaluation of computer programs used for structural analyses of impact response of spent fuel shipping casks

    International Nuclear Information System (INIS)

    This report presents the results of a study of impact analyses of a generic spent-fuel cask. The study compares the use and results of three different finite element computer codes. Seven different cask-like model analyses are considered. The models encompass both linear and nonlinear geometric and material behavior. On the basis of the analyses results, this report recommends what parameters are useful in the comparison of different structural finite element computer programs. 5 references, 36 figures, 11 tables

  8. Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    International Nuclear Information System (INIS)

    This Preliminary Design Report (PDR) provides a detailed description of the design, analyses, and testing programs for the BR-100 cask. The BR-100 is a Type B(U) cask designed for transport by rail or barge. This report presents the preliminary analyses and tests which have been performed for the BR-100 and outlines the confirmatory analyses and tests which will be performed

  9. Status of cask procurement strategy to satisfy DOE/OCRWM requirements

    International Nuclear Information System (INIS)

    The Nuclear Waste Policy Act requires the development of a safe and efficient system to transport spent nuclear fuel to and within the Federal Waste Management System. This paper describes the DOE/OCRWM strategy to develop and procure a major component of the Transportation System-the transport cask systems. The original initiative to develop high-capacity innovative designs and its current status is described. The follow-on phase to design and procure proven technology cask systems is also discussed

  10. Final report on shipping-cask sabotage source-term investigation

    International Nuclear Information System (INIS)

    A need existed to estimate the source term resulting from a sabotage attack on a spent nuclear fuel shipping cask. An experimental program sponsored by the US NRC and conducted at Battelle's Columbus Laboratories was designed to meet that need. In the program a precision shaped charge was fired through a subscale model cask loaded with segments of spent PWR fuel rods and the radioactive material released was analyzed. This report describes these experiments and presents their results

  11. Spent fuel shipping cask handling capability assessment of 27 selected light water reactors

    International Nuclear Information System (INIS)

    This report presents an assessment of the spent fuel shipping cask handling capabilities of those nuclear plants currently projected to lose full core reserve capability in their spent fuel storage basins in the near future. The purpose of this assessment is to determine which cask types, in the current fleet, each of the selected reactors can handle. The cask handling capability of a nuclear plant depends upon both external and internal conditions at the plant. The availability of a rail spur, the lifting capacity of the crane, the adequacy of clearances in the cask receiving, loading, and decontamination areas and similar factors can limit the types of casks that can be utilized at a particular plant. This report addresses the major facility capabilities used in assessing the types of spent fuel shipping casks that can be handled at each of the 27 selected nuclear plants approaching a critical storage situation. The results of this study cannot be considered to be final and are not intended to be used to force utilities to ship by a particular mode. In addition, many utilities have never shipped spent fuel. Readers are cautioned that the results of this study reflect the current situation at the selected plants and are based on operator perceptions and guidance from NRC related to the control of heavy loads at nuclear power plants. Thus, the cask handling capabilities essentially represent snap-shots in time and could be subject to change as plants further analyze their capabilities, even in the near-term. The results of this assessment indicate that 48% of the selected plants have rail access and 59% are judged to be candidates for overweight truck shipments (with 8 unknowns due to unavailability of verifiable data). Essentially all of the reactors can accommodate existing legal-weight truck casks. 12 references, 1 figure, 4 tables

  12. Evaluation of computer programs used for structural analyses of impact response of spent fuel shipping casks

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B A; Gwinn, K W

    1984-05-01

    This report presents the results of a study of impact analyses of a generic spent-fuel cask. The study compares the use and results of three different finite element computer codes. Seven different cask-like model analyses are considered. The models encompass both linear and nonlinear geometric and material behavior. On the basis of the analyses results, this report recommends what parameters are useful in the comparison of different structural finite element computer programs. 5 references, 36 figures, 11 tables.

  13. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, Kaushik [ORNL; Scaglione, John M [ORNL

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 Δkeff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  14. Operations manual for the Beneficial Uses Shipping System cask. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Bronowski, D.R.; Yoshimura, H.R.

    1993-04-01

    This document is the Operations Manual for the Beneficial Uses Shipping System (BUSS) cask. These operating instructions address requirements; for loading, shipping, and unloading, supplementing general operational information found in the BUSS Safety Analysis Report for Packaging (SARP), SAND 83-0698. Use of the BUSS cask is authorized by Department of Energy (DOE) and Nuclear Regulatory Commission (NRC) for the shipment of special form cesium chloride or strontium flouride capsules.

  15. SAS1 and SAS4, two new shielding analysis sequences for spent fuel casks

    International Nuclear Information System (INIS)

    Two important Shielding Analysis Sequences (SAS) have recently been developed within the SCALE computational system. These sequences significantly enhance the existing SCALE system capabilities for evaluating radiation doses exterior to spent fuel casks. These new control module sequences (SAS1 and SAS4) and their capabilities are discussed and demonstrated. Particular attention is given to the new SAS4 sequence which provides an automated scheme for generating and using biasing parameters in a subsequent Monte Carlo analysis of a cask

  16. COBRA-SFS modifications and cask model optimization

    Energy Technology Data Exchange (ETDEWEB)

    Rector, D.R.; Michener, T.E.

    1989-01-01

    Spent-fuel storage systems are complex systems and developing a computational model for one can be a difficult task. The COBRA-SFS computer code provides many capabilities for modeling the details of these systems, but these capabilities can also allow users to specify a more complex model than necessary. This report provides important guidance to users that dramatically reduces the size of the model while maintaining the accuracy of the calculation. A series of model optimization studies was performed, based on the TN-24P spent-fuel storage cask, to determine the optimal model geometry. Expanded modeling capabilities of the code are also described. These include adding fluid shear stress terms and a detailed plenum model. The mathematical models for each code modification are described, along with the associated verification results. 22 refs., 107 figs., 7 tabs.

  17. Marginal overweight operating scenario for DOE's initiative I highway casks

    International Nuclear Information System (INIS)

    This paper assesses the potential transport of high-capacity Initiative I highway casks under development by the Office of Civilian Radioactive Waste Management (OCRWM) as permitted marginal overweight shipments that: exceed a gross vehicle weight (gvw) limit of 80,000, but weight less than 96,000 pounds; follow axle and axle group weight limits adopted by the Surface Transportation Assistance Act (STAA) of 1982; conform to dimensional restrictions to operate on most major highways; and comply with the Federal Bridge Formula. The marginal overweight tractor-trailer would operate in normal open-quotes over-the-roadclose quotes mode and comply with all laws and regulations. The vehicle would have a sleeper berth and two drivers - one to drive while the other provides escort and communications services and accumulates required off-duty time

  18. Kajian Bentuk Pengolahan dan Analisis Finansial Buah Api api (Avicennia officinalis L.) Sebagai Bahan Makanan dan Minuman di Kabupaten Deli Serdang

    OpenAIRE

    Sianturi, Gustinaria

    2013-01-01

    Penelitian ini bertujuan untuk mengidentifikasi bentuk pengolahan, tingkat kelayakan nilai finansial, dan strategi pengembangan usaha pengolahan buah api-api (Avicennia officinalis L.) di Dusun Paluh Merbau, Desa Tanjung Rejo, Kecamatan Percut Sei Tuan, Kabupaten Deli Serdang. Metode analisis yang digunakan adalah analisis deskriptif, analisis finansial, dan analisis strategi pengembangan usaha dengan menggunakan analisis SWOT. Hasil penelitian menunjukkan bahwa produk olahan buah api-...

  19. Analisis Faktor yang Mempengaruhi Indeks Prestasi Mahasiswa

    Directory of Open Access Journals (Sweden)

    Putriaji Hendikawati

    2011-12-01

    Full Text Available Penelitian ini dilaksanakan untuk mengungkap dan menganalisis faktor-faktor yangmempengaruhi perolehan indeks prestasi mahasiswa. Populasi penelitian adalahmahasiswa program studi Pendidikan Matematika FMIPA Unnes dan dipilih sampelsebanyak 3 kelas. Pengambilan sampel dilakukan dengan cara stratified cluster randomsamplingdengan sampel penelitian berjumlah 114 mahasiswa.Hasil penelitian menunjukkan bahwa Indeks Prestasi (IP mahasiswa dipengaruhioleh beberapa variabel antara lain: variabel suasana hati, membagi waktu, hubungandengan keluarga, penjelasan dosen, suasana tempat tinggal, kegiatan selain kuliah, bakat,adaptasi lingkungan, pantauan orang tua, perhatian orang tua, pergaulan, makan dan gizi,IQ dan EQ, kemampuan sosialisasi, kondisi keuangan, suasana belajar kampus, pancaindera kemampuan menangkap materi, dan olahraga. Setelahdilakukan analisis faktor danproses reduksi diperoleh5 faktor yang mempengaruhi IP mahasiswa. Lima faktor tersebutadalah Faktor Manajemen Diri, Faktor Lingkungan Sekitar, Faktor Kondisi Eksternal,Faktor Kondisi Fisik dan Faktor Olahraga.Hasil penelitian ini bermanfaat bagi mahasiswa serta para dosen khususnyapenentu kebijakan di jurusan Matematika FMIPA Unnes, untuk mengembangkan sertameningkatkan faktor-faktor yang mempengaruhi IP mahasiswa yang berhubungan dengankebijakan dalam kampus agar dapat memberikan kontribusi positif bagi perolehan IPmahasiswa. Kata kunci: analisis faktor, indeks prestasi, mahasiswa.

  20. SCANS (Shipping Cask ANalysis System) a microcomputer-based analysis system for shipping cask design review: User's manual to Version 3a. Volume 1, Revision 2

    International Nuclear Information System (INIS)

    SCANS (Shipping Cask ANalysis System) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for evaluating safety analysis reports on spent fuel shipping casks. SCANS is an easy-to-use system that calculates the global response to impact loads, pressure loads and thermal conditions, providing reviewers with an independent check on analyses submitted by licensees. SCANS is based on microcomputers compatible with the IBM-PC family of computers. The system is composed of a series of menus, input programs, cask analysis programs, and output display programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests. Analysis options are based on regulatory cases described in the Code of Federal Regulations 10 CFR 71 and Regulatory Guides published by the US Nuclear Regulatory Commission in 1977 and 1978

  1. SCANS (Shipping Cask ANalysis System) a microcomputer-based analysis system for shipping cask design review: User`s manual to Version 3a. Volume 1, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Mok, G.C.; Thomas, G.R.; Gerhard, M.A.; Trummer, D.J.; Johnson, G.L. [Lawrence Livermore National Lab., CA (United States)

    1998-03-01

    SCANS (Shipping Cask ANalysis System) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for evaluating safety analysis reports on spent fuel shipping casks. SCANS is an easy-to-use system that calculates the global response to impact loads, pressure loads and thermal conditions, providing reviewers with an independent check on analyses submitted by licensees. SCANS is based on microcomputers compatible with the IBM-PC family of computers. The system is composed of a series of menus, input programs, cask analysis programs, and output display programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests. Analysis options are based on regulatory cases described in the Code of Federal Regulations 10 CFR 71 and Regulatory Guides published by the US Nuclear Regulatory Commission in 1977 and 1978.

  2. Estimation of terrorist attack resistibility of dual-purpose cask TP-117 with DU (depleted uranium) gamma shield

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, O.G.; Matveev, V.Z.; Morenko, A.I.; Il' kaev, R.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation)

    2004-07-01

    Report is devoted to numerical research of dual-purpose unified cask (used for SFA transportation and storage) resistance to terrorist attacks. High resistance of dual-purpose unified cask has been achieved due to the unique design-technological solutions and implementation of depleted uranium in cask construction. In suggested variant of construction depleted uranium fulfils functions of shielding and constructional material. It is used both in metallic and cermet form (basing on steel and depleted uranium dioxide). Implementation of depleted uranium in cask construction allows maximal load in existing overall dimensions of the cask. At the same time: 1) all safety requirements (IAEA) are met, 2) dual-purpose cask with SFA has high resistance to terrorist attacks.

  3. Estimation of terrorist attack resistibility of dual-purpose cask TP-117 with DU (depleted uranium) gamma shield

    International Nuclear Information System (INIS)

    Report is devoted to numerical research of dual-purpose unified cask (used for SFA transportation and storage) resistance to terrorist attacks. High resistance of dual-purpose unified cask has been achieved due to the unique design-technological solutions and implementation of depleted uranium in cask construction. In suggested variant of construction depleted uranium fulfils functions of shielding and constructional material. It is used both in metallic and cermet form (basing on steel and depleted uranium dioxide). Implementation of depleted uranium in cask construction allows maximal load in existing overall dimensions of the cask. At the same time: 1) all safety requirements (IAEA) are met, 2) dual-purpose cask with SFA has high resistance to terrorist attacks

  4. Use of transportable storage casks in the nuclear waste management system

    International Nuclear Information System (INIS)

    A study was performed to determine the viability of the use of transportable storage casks (TSCs), and other metal casks that are designed primarily for storage but which might be used to ship their stored contents to DOE on a one-time use basis (referred to in this study as storage only casks, or SOCs), in the combined utility/DOE spent fuel management system. The viability of the use of TSCs and SOCs was assessed in terms of the costs and savings involved in their use, the sensitivity of these costs and savings to changes in the capacity and cost of fabrication of the casks, the impacts of variation in cask design features on cost and radiation exposure of personnel, and their prospective use in connection with the transport of defense high level wastes. Estimates were developed of the costs of acquiring and handling of TSCs and SOCs at reactor sites. For comparison purposes, similar costs were developed for the use of concrete storage casks at reactor sites. Estimates of the savings involved to the DOE system as a result of receiving spent fuel in TSCs or SOCs were separately developed. A summary of the results of cost estimates developed in Section 4.0 and the Appendices to this report is shown in Tables 2-1 and 2-2 for instances in which the TSC or SOC were delivered to DOE containing intact fuel assemblies and cans of consolidated fuel, respectively. 2 figs., 14 tabs

  5. Use of transportable storage casks in the nuclear waste management system: Appendices

    International Nuclear Information System (INIS)

    A study was performed to determine the viability of the use of transportable storage casks (TSCs), and other metal casks that are designed primarily for storage but which might be used to ship their stored contents to DOE on a one-time use basis (referred to in this study as storage only casks, or SOCs), in the combined utility/DOE spent fuel management system. The viability of the use of TSCs and SOCs was assessed in terms of the costs and savings involved in their use, the sensitivity of these costs and savings to changes in the capacity and cost of fabrication of the casks, the impacts of variation in cask design features on cost and radiation exposure of personnel, and their prospective use in connection with the transport of defense high level wastes. Estimates were developed of the costs of acquiring and handling of TSCs and SOCs at reactor sites. For comparison purposes, similar costs were developed for the use of concrete storage casks at reactor sites. Estimates of the savings involved to the DOE system as a result of receiving spent fuel in TSCs or SOCs were separately developed. These costs are developed and presented in Volume 2, Appendices A through J

  6. Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

    Energy Technology Data Exchange (ETDEWEB)

    Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge Northridge, CA 91330 (United States)

    2012-07-01

    Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National

  7. Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

    International Nuclear Information System (INIS)

    Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's

  8. Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2002-12-17

    This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).

  9. PENENTUAN DAERAH RAWAN GIZI BERDASARKAN ANALISIS SPATIAL

    Directory of Open Access Journals (Sweden)

    Noviati Fuada

    2012-11-01

    Full Text Available Latar Belakang : Riset Kesehatan Dasar telah dilakukan di Indonesia (RISKESDAS 2007. Riset telah mengumpulkan data-data yang terdiri dari data kesehatan yang menggambarkan status gizi anak di bawah lima (antrophometri data di seluruh wilayah Indonesia. Kenyataanya masih sedikit analisis dengan menggunakan metode GIS, oleh karena itu artikel ini akan dikaji dengan metode spasial. Kajian ini diharapkan dapat memberikan informasi faktual, yang dapat mendukung kebijakan daerah. Tujuan: Mengidentifikasi daerah kabupaten/provinsi rawan status gizi anak balita, Metode: Analisa GIS denganmenggunakan metode spasial (pengelompokan data dan overlay dengan cara union. Data RISKESDAS 2007. Hasil: Wilayah tingkat tinggi potensi rawan gizi bermasalah (bersumber overlay antara peta sebaran status gizi balita dengan peta sebaran KK miskin  adalah; Kota Tasikmalaya, Kab. Tasikmalaya, Cianjur, Garut, Ciamis, Bandung, Subang dan Majalengka. Wilayah tingkat tinggi berpotensi terkena infeksi penyakit (berdasarkan peta sebaran resiko Infeksi Penyakit dan pemanfaatan posyandu adalah: Kabupaten Purwakarta, Karawang, Bekasi, Bogor, Sukabumi, Tasikmalaya, Kota Tasikmalaya, Bekasi dan Bogor. Wilayah berpotensi rawan gizi kategori tinggi (bersumber pada 4 faktor/peta sebaran meliputi, Kabupaten Cianjur, Garut, Tasikmalaya dan Kota Tasikmalaya. Kasus Gizi bermasalah berdasarkan 3 indeks gabungan menyebar di seluruh wilayah Provinsi Jawa Barat. Wilayah kasus gizi bermasalah kategori tinggi, dan kategori sedang, sebagian besar  terjadi di wilayah Kabupaten. Baik kategori sedang maupun tinggi merupakan wilayah yang berdampingan. Gambaran ini mengarah pada fakta bahwa  masalah gizi cenderung merupakan masalah epidemiologi. Kesimpulan: Terdapat empat wilayah kabupaten status gizi yang paling serius dalam kategori tinggi meliputi, Kabupaten Cianjur, Garut, Tasikmalaya dan Kota Tasikmalaya.  Kata kunci: analisis spasial, status gizi, posyandu, rawan gizi 

  10. ANALISIS VARIABEL KEUANGAN YANG MEMPENGARUHI KEBIJAKAN DEVIDEN

    Directory of Open Access Journals (Sweden)

    Sumiadji -

    2012-03-01

    Full Text Available Penelitian ini dilakukan untuk menguji pengaruh variabel keuangan yang terdiri dari: rasio pro-fitabilitas (return on assets, likuiditas (cash ratio, rasio hutang (debt to equity ratio, market value (earnings per share, dan perputaran total aset (total assets turnover terhadap kebijakan dividen yang diproksikan dengan dividend payout ratio (DPR pada perusahaan manufaktur yang terdaftar di Bursa Efek Indonesia tahun 2004-2008. Prosedur pemilihan sampel penelitian menggunakan purposive sampling sehingga menghasilkan 8 perusahaan yang memenuhi kriteria sampel. Data sekunder dikumpulkan dengan teknik dokumentasi bersumber dari Indonesian Capital Market Directory (ICMD, laporan keuangan, dan hasil Rapat Umum Pemegang Saham (RUPS. Teknik analisis yang digunakan dalam penelitian ini adalah analisis regresi linear berganda. Hasil penelitian ini adalah secara simultan variabel ROA, CR, DER, EPS dan TATO berpengaruh terhadap DPR. Secara parsial variabel yang mempengaruhi DPR adalah CR, EPS dan TATO. Varabel lainnya, yaitu ROA dan DER ditemukan tidak berpengaruh terhadap DPR. This research was conducted to examine the influence of the financial variables which consists of: return on assets (ROA, cash ratio (CR, debt to equity ratio (DER, earnings per share (EPS, and total asset turnover (TATO to the dividend policy that indicated by the dividend payout ratio (DPR of listed manufacturing company in Indonesia Stock Exchange  2004 to 2008. The sample selection procedure used was a purposive sampling so that it produced eight companies that met the sample criteria. Se-condary data was collected by the documentation technique were obtained from the Indonesian Capital Market Directory (ICMD, financial statements, and the results of the Annual General Meeting of Shareholders. The analysis technique used was multiple linear regression analysis. The research found that five variables of kind of ROA, CR, DER, EPS dan TATO simultaneously influence to dividend

  11. Conceptual Design Report Cask Loadout Sys and Cask Drop Redesign for the Immersion Pail Support Structure and Operator Interface Platform at 105 K West

    Energy Technology Data Exchange (ETDEWEB)

    LANGEVIN, A.S.

    1999-07-12

    This conceptual design report documents the redesign of the IPSS and the OIP in the 105 KW Basin south loadout pit due to a postulated cask drop accident, as part of Project A.5/A.6, Canister Transfer Facility Modifications. Project A.5/A.6 involves facility modifications needed to transfer fuel from the basin into the cask-MCO. The function of the IPSS is to suspend, guide, and position the immersion pail. The immersion pail protects the cask-MCO from contamination by basin water and acts as a lifting device for the cask-MCO. The OIP provides operator access to the south loadout pit. Previous analyses studied the effects of a cask-MCO drop on the south loadout pit concrete structure and on the IPSS. The most recent analysis considered the resulting loads at the pit slab/wall joint (Kanjilal, 1999). This area had not been modeled previously, and the analysis results indicate that the demand capacity exceeds the allowable at the slab/wall joint. The energy induced on the south loadout pit must be limited such that the safety class function of the basin is maintained. The solution presented in this CDR redesigns the IPSS and the OIP to include impact-absorbing features that will reduce the induced energy. The impact absorbing features of the new design include: Impact-absorbing material at the IPSS base and at the upper portion of the IPSS legs. A sleeve which provides a hydraulic means of absorbing energy. Designing the OIP to act as an impact absorber. The existing IPSS structure in 105 KW will be removed. This conceptual design considers only loads resulting from drops directly over the IPSS and south loadout pit area. Drops in other areas of the basin are not considered, and will be covered as part of a future revision to this CDR.

  12. Conceptual Design Report - Cask Loadout System Cask Drop Redesign for the Immersion Pail Support Structure and Operator Interface Platform, 105 K West Basin, Project A.5/A.6

    International Nuclear Information System (INIS)

    This conceptual design report documents the redesign of the immersion pail support structure (IPSS) and the OIP in the 105 KW Basin south loadout pit due to a postulated cask drop accident, as part of Project A.5lA.6, Canister Transfer Facility Modifications. Project A.5lA.6 involves facility modifications needed to transfer fuel from the basin into the cask-MCO. The junction of the IPSS is to suspend, guide, and position the immersion pail. The immersion pail protects the cask-MCO from contamination by basin water and acts as a lifting device for the cask-MCO. The OIP provides operator access to the south loadout pit. Previous analyses studied the effects of a cask-MCO drop on the south loadout pit concrete structure and on the IPSS. The most recent analysis considered the resulting loads at the pit slab/wall joint (Kanjilal, 1999). This area had not been modeled previously, and the analysis results indicate that the demand capacity exceeds the allowable at the slablwall joint. The energy induced on the south loadout pit must be limited such that the safety class function of the basin is maintained. The solution presented in this CDR redesigns the IPSS and the OIP to include impact-absorbing features that will reduce the induced energy. The impact absorbing features of the new design include: Impact-absorbing material at the IPSS base and at the upper portion of the IPSS legs. A sleeve which provides a hydraulic means of absorbing energy. Designing the OIP to act as an impact absorber. The existing IPSS structure in 105 KW will be removed. This conceptual design considers only loads resulting from drops directly over the IPSS and south loadout pit area. Drops in other areas of the basin are not considered, and will be covered as part of a future revision to this CDR

  13. Documentation for fiscal year 1995 annual BUSS cask SARP testing and inspections

    Energy Technology Data Exchange (ETDEWEB)

    Saueressig, P.T.

    1994-11-08

    The purpose of this report is to compile the data generated during the Fiscal Year (FY) 1995 annual tests and inspections performed on the Beneficial Uses Shipping System (BUSS) cask. The BUSS Cask Model R-1 is a type B shipping container used for shipment of radioactive cesium-137 and strontium-90 capsules to Waste Encapsulation and Storage Facility (WESF). The primary purpose of the BUSS Cask is to provide shielding and confinement as well as impact, puncture, and thermal protection for the capsules under both normal and accident conditions. Section 8.2 ``Maintenance and Periodic Inspection Program`` of the BUSS Cask SARP requires that the following tests and inspections be performed on an annual basis: hydrostatic pressure test; helium leak test; dye penetrant test on the trunnions and life lugs; torque test on all permanent bolts; and impact limiter inspection and weight test. In addition to compiling the generated data, this report will verify that the testing criteria identified in section 8.2 of the BUSS Cask Safety Analysis Report for Packaging (SARP) was met.

  14. Castor-1C spent fuel storage cask decay heat, heat transfer, and shielding analyses

    International Nuclear Information System (INIS)

    This report documents the decay heat, heat transfer, and shielding analyses of the Gesellschaft fuer Nuklear Services (GNS) CASTOR-1C cask used in a spent fuel storage demonstration performed at Preussen Elektra's Wurgassen nuclear power plant. The demonstration was performed between March 1982 and January 1984, and resulted in cask and fuel temperature data and cask exterior surface gamma-ray and neutron radiation dose rate measurements. The purpose of the analyses reported here was to evaluate decay heat, heat transfer, and shielding computer codes. The analyses consisted of (1) performing pre-look predictions (predictions performed before the analysts were provided the test data), (2) comparing ORIGEN2 (decay heat), COBRA-SFS and HYDRA (heat transfer), and QAD and DOT (shielding) results to data, and (3) performing post-test analyses if appropriate. Even though two heat transfer codes were used to predict CASTOR-1C cask test data, no attempt was made to compare the two codes. The codes are being evaluated with other test data (single-assembly data and other cask data), and to compare the codes based on one set of data may be premature and lead to erroneous conclusions

  15. Structural evaluation and analysis under normal conditions for spent fuel concrete storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Taechul; Baeg, Changyeal; Yoon, Sitae [Korea Radioactive waste Management Agency, Daejeon (Korea, Republic of); Jung, Insoo [Korea Nuclear Engineering and Service Co., Daejeon (Korea, Republic of)

    2014-05-15

    The purpose of this paper is the verification of stabilities of the structural elements that influence the safety of a concrete storage cask. The evaluation results were reviewed with respect to every design criterion, in terms of whether the results satisfy the criteria, provided by 10CFR 72 and NUREG-1536. The basic information on the design is partially explained in 2. Description of spent fuel storage system and the maintainability and assumptions included in the analysis were confirmed through detailed explanations of the acceptable standards, analysis model, and analysis method. ABAQUS 6.10, a widely used finite element analysis program, was used in the structural analysis. The storage cask shall maintain the sub-criticality, shielding, structural integrity, thermal capability and confinement in accordance with the requirements specified in US 10 CFR 72. The safety of storage cask is analyzed and it has been confirmed to meet the requirements of US 10 CFR 72. This paper summarizes the structural stability evaluation results of a concrete storage cask with respect to the design criteria. The evaluation results of this paper show that the maximum stress was below the allowable stress under every condition, and the concrete storage cask satisfied the design criteria.

  16. Documentation for fiscal year 1995 annual BUSS cask SARP testing and inspections

    International Nuclear Information System (INIS)

    The purpose of this report is to compile the data generated during the Fiscal Year (FY) 1995 annual tests and inspections performed on the Beneficial Uses Shipping System (BUSS) cask. The BUSS Cask Model R-1 is a type B shipping container used for shipment of radioactive cesium-137 and strontium-90 capsules to Waste Encapsulation and Storage Facility (WESF). The primary purpose of the BUSS Cask is to provide shielding and confinement as well as impact, puncture, and thermal protection for the capsules under both normal and accident conditions. Section 8.2 ''Maintenance and Periodic Inspection Program'' of the BUSS Cask SARP requires that the following tests and inspections be performed on an annual basis: hydrostatic pressure test; helium leak test; dye penetrant test on the trunnions and life lugs; torque test on all permanent bolts; and impact limiter inspection and weight test. In addition to compiling the generated data, this report will verify that the testing criteria identified in section 8.2 of the BUSS Cask Safety Analysis Report for Packaging (SARP) was met

  17. Structural evaluation and analysis under normal conditions for spent fuel concrete storage cask

    International Nuclear Information System (INIS)

    The purpose of this paper is the verification of stabilities of the structural elements that influence the safety of a concrete storage cask. The evaluation results were reviewed with respect to every design criterion, in terms of whether the results satisfy the criteria, provided by 10CFR 72 and NUREG-1536. The basic information on the design is partially explained in 2. Description of spent fuel storage system and the maintainability and assumptions included in the analysis were confirmed through detailed explanations of the acceptable standards, analysis model, and analysis method. ABAQUS 6.10, a widely used finite element analysis program, was used in the structural analysis. The storage cask shall maintain the sub-criticality, shielding, structural integrity, thermal capability and confinement in accordance with the requirements specified in US 10 CFR 72. The safety of storage cask is analyzed and it has been confirmed to meet the requirements of US 10 CFR 72. This paper summarizes the structural stability evaluation results of a concrete storage cask with respect to the design criteria. The evaluation results of this paper show that the maximum stress was below the allowable stress under every condition, and the concrete storage cask satisfied the design criteria

  18. Integrated cask storage systems for storage, transportation, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Since 1979 Tennessee Valley Authority TVA has participated in conceptual design studies of dry storage vaults, silos, casks, ad dry wells, and, with DOE and others, has undertaken limited demonstrations of rod consolidation and cask dry storage at TVA's Browns Ferry Nuclear Plant in Alabama. TVA believes the integrated storage cask concept is worthy of consideration as an alternative for spent fuel management. Placing spent fuel in a secure passive storage mode at an early date and avoiding unnecessary handling and repackaging reduces the potential for occupational and public radiological exposure. Therefore the notion of a universal cask used for storage, shipment, and disposal is appealing from a safety, environmental, and public perception standpoint. The universal cask can also serve as a dispersed monitored retrievable storage (MRS), thus eliminating the need for redundant facilities, and it does not foreclose future options. It also appears that this concept would simplify repository design, ease retrievability, and provide greater flexibility in repository siting. 2 figures, 2 tables

  19. An attempt to estimate gamma-ray dose rate from radioactive waste storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M.; Plecas, I.; Pavlovic, R.; Pavlovic, S. [Institute for Nuclear Sciences ' ' Vinca' ' , Belgrade (Yugoslavia); Sokcic-Kostic, M. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Hauptabteilung Zyklotron

    2001-07-01

    Radioactive waste - {sup 137}Cs and {sup 60}Co contaminated sludge from the irradiated fuel storage pool of the RA research reactor in the Vinca Institute of Nuclear Sciences, Belgrade, is conditioned and stored in specially designed casks in 1997. Main purpose of this paper is to describe an attempt to estimate a gamma-ray ambient dose equivalent rate from the cask with the conditioned sludge by reference Monte Carlo MCNP code and compare the result to the measuring data. The aim of the study is to master with a reliable computational tool that allows faithful estimation of the total ambient gamma-ray dose equivalent rates from the cask. Values of so obtained gamma-ray ambient dose equivalent rates are compared to the measured values at the same spots of the cask and acceptable agreement were found. These data could be used in a further study on minimisation of the total ambient dose equivalent rate of a regular array or random pile of casks, stored in the storage location. (orig.)

  20. Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance.

  1. Safety analysis report vitrified high level waste type B shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    This Safety Analysis Report describes the design, analyses, and principle features of the Vitrified High Level Waste (VHLW) Cask. In preparing this report a detailed evaluation of the design has been performed to ensure that all safety, licensing, and operational goals for the cask and its associated Department of Energy program can be met. The functions of this report are: (1) to fully document that all functional and regulatory requirements of 10CFR71 can be met by the package; and (2) to document the design and analyses of the cask for review by the Nuclear Regulatory Commission. The VHLW Cask is the reusable shipping package designed by GNSI under Department of Energy contract DE-AC04-89AL53-689 for transportation of Vitrified High Level Waste, and to meet the requirements for certification under 10CFR71 for a Type B(U) package. The VHLW cask has been designed as packaging for transport of canisters of Vitrified High Level Waste solidified at Department of Energy facilities.

  2. Behavior of Full-Scale Model Cask Under 9 m Drop Test and Simulation

    International Nuclear Information System (INIS)

    The nuclear spent fuel transport and storage cask is used for transport of the spent fuel from a nuclear power station to an intermediate storage facility. Leak tightness and subcriticality on transportation required from IAEA TS-R1[1] have to be assured by a 9 m drop test and its numerical simulation. This paper describes the drop test using a full-scale prototype test cask The test was conducted by German Federal Institute for Materials Research and Testing (BAM) at their test facility in Horstwalde, Germany and comparison of the test result with the 'MH1 (Mitsubishi Heavy Industries, Ltd.)' numerical simulation using LS-DYNA code. The drop orientations of the tests were slap down and vertical. From the drop test the following is demonstrated: The leak rate of He gas after the drop tests satisfied the IAEA's criteria. The numerical simulation which modeled the cask body enabled dynamic response such as acceleration and strain of the cask body. This means the simulation method qualified the relation of dynamic response of the cask body and leakage behavior. (authors)

  3. Evaluation of Impact Resistance of Concrete Overpack of Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sanghoon; Kim, Ki-Young; Jeon, Je-Eon; Seo, Ki-Seog [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The concrete overpack of the cask provides radiation shielding as well as physical protection for inner canister against external mechanical shock. When the overpack undergoes a severe missile impact which might be caused by tornado or aircraft crash, it should sustain minimal level of structural integrity so that the radiation shielding and the retrievability of canister are maintained. Empirical formulas have been developed for the evaluation of concrete damage but those formulas can be used only for local damage evaluation and not for global damage evaluation. In this research, a series of numerical simulations and tests have been performed to evaluate the damage of two types of concrete overpack segment models under high speed missile impact. It is shown that appropriate modeling of material failure is crucial in this kind of analyses and finding the correct failure parameters may not be straightforward. When comparing the simulation results with the test results, it is shown that neither setting, case 1 and 2 provides results with consistent agreement with test results. That is, case 1 setting is more close to reality in Type 1 model analysis, but for Type 2, case 2 setting provides more close results to the reality. In both the case, not enough deformation is predicted by simulation compared to the tests. Weak failure and eroding criteria give larger penetration depth with insufficient overall damage due to energy loss with element erosion.

  4. An analysis of contingencies for making casks available for use during the early years of federal waste management system operations

    International Nuclear Information System (INIS)

    This paper reports on a study which has been performed to examine the contingencies that could be pursued by the Department of energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) for shipping spent fuel beginning in 1998. OCRWM's current plan is to initiate operations using early production units of Initiative I truck and rail/barge casks that are presently being designed. Contingencies to this plan were considered in case some unforeseen event occurs that precludes the Initiative I casks from entering into service early in 1998 in sufficient quantities (both numbers and types) to satisfy DOE's shipping needs. Specifically, the study addressed the potential availability of cask systems, selected several cask usage scenarios, determined the requirements for casks under these scenarios, generically assessed different strategies for acquiring casks or the use of casks, and generically assessed cask fabrication capabilities. Issues concerning both domestic and foreign resources were addressed with a focus on the first five years of Federal Waste Management System (FWMS) operation

  5. A methodology for estimating the residual contamination contribution to the source term in a spent-fuel transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, T.L. (Sandia National Labs., Albuquerque, NM (United States)); Jordan, H. (EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant); Pasupathi, V. (Battelle, Columbus, OH (United States)); Mings, W.J. (USDOE Idaho Field Office, Idaho Falls, ID (United States)); Reardon, P.C. (GRAM, Inc., Albuquerque, NM (United States))

    1991-09-01

    This report describes the ranges of the residual contamination that may build up in spent-fuel transport casks. These contamination ranges are calculated based on data taken from published reports and from previously unpublished data supplied by cask transporters. The data involve dose rate measurements, interior smear surveys, and analyses of water flushed out of cask cavities during decontamination operations. A methodology has been developed to estimate the effect of residual contamination on spent-fuel cask containment requirements. Factors in estimating the maximum permissible leak rates include the form of the residual contamination; possible release modes; internal gas-borne depletion; and the temperature, pressure, and vibration characteristics of the cask during transport under normal and accident conditions. 12 refs., 9 figs., 4 tabs.

  6. A methodology for estimating the residual contamination contribution to the source term in a spent-fuel transport cask

    International Nuclear Information System (INIS)

    This report describes the ranges of the residual contamination that may build up in spent-fuel transport casks. These contamination ranges are calculated based on data taken from published reports and from previously unpublished data supplied by cask transporters. The data involve dose rate measurements, interior smear surveys, and analyses of water flushed out of cask cavities during decontamination operations. A methodology has been developed to estimate the effect of residual contamination on spent-fuel cask containment requirements. Factors in estimating the maximum permissible leak rates include the form of the residual contamination; possible release modes; internal gas-borne depletion; and the temperature, pressure, and vibration characteristics of the cask during transport under normal and accident conditions. 12 refs., 9 figs., 4 tabs

  7. Neutron measurements around storage casks containing spent fuel and vitrified high-level radioactive waste at ZWILAG.

    Science.gov (United States)

    Buchillier, T; Aroua, A; Bochud, F O

    2007-01-01

    Spectrometric and dosimetric measurements were made around a cask containing spent fuel and a cask containing high-level radioactive waste at the Swiss intermediate waste and spent fuel storage facility. A Bonner sphere spectrometer, an LB 6411 neutron monitor and an Automess Szintomat 6134A were used to characterise the n-gamma fields at several locations around the two casks. The results of these measurements show that the neutron fluence spectra around the cask containing radioactive waste are harder and higher in intensity than those measured in the vicinity of the spent fuel cask. The ambient dose equivalents measured with the LB 6411 neutron monitor are in good agreement with those obtained using the Bonner spheres, except for locations with soft neutron spectra where the monitor overestimates the neutron ambient dose equivalent by almost 50%. PMID:17494980

  8. Structural design of concrete storage pads for spent-fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Y.R.; Nickell, R.E.; James, R.J. (ANATECH Research Corp., San Diego, CA (United States))

    1993-04-01

    The loading experienced by spent fuel dry storage casks and storage pads due to potential drop or tip-over accidents is evaluated using state-of-the-art concrete structural analysis methodology. The purpose of this analysis is to provide simple design charts and formulas so that design adequacy of storage pads and dry storage casks can be demonstrated. The analysis covers a wide range of slab-design parameters, e.g., reinforcement ratio, slab thickness, concrete compressive strength, and sub-base soil compaction, as well as variations in drop orientation and drop height. The results are presented in the form of curves, giving the force on the cask as a function of storage pad hardness for various drop heights. In addition, force-displacement curves, deformed shapes, crack patterns, stresses and strains are given for various slab-design conditions and drop events. The utility of the results in design are illustrated through examples.

  9. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    Science.gov (United States)

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. PMID:26720262

  10. ITER Upper Port Plug handling cask system assessment and design proposals

    International Nuclear Information System (INIS)

    The current design of the ITER cask for Upper Port Plugs has been evaluated. Careful reduction of the number of mechanical degrees of freedom is an opportunity to relax the tolerances in the design, resulting in cost reduction and reliability increase. A new kinematical design for the tractor module has a higher stiffness to weight ratio, reduces actuator forces by a factor four and minimizes cross-talk between lift and rotation motion. Non-cantilevered handling is recommended to reduce wheel loads on the tractor by a factor six and to simplify guidance. At the system level the tubular guide (TG) is proposed, a semi-permanent 3.5 m long tube which is an extension of the Upper Port. Cask docking is simplified and the risk of the cask tilting is prevented. Redesigning the system concept is recommended and the TG looks promising. Since a system level redesign impacts the external interfaces, overall feasibility has to be investigated.

  11. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    International Nuclear Information System (INIS)

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  12. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Jang, Jung Nam; Hwang, Yong Hwa; Kwon, In Chan; Min, Duck Kee; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  13. Ultrasonic inspection techniques for two weld closures proposed for RSSF waste storage casks

    International Nuclear Information System (INIS)

    One method being considered for interim storage of high-level radioactive waste materials is to place these materials in large sealed stainless steel canisters and subsequently store these canisters in a second sealed steel storage cask. Weld procedures are proposed as the closure or seal for these vessels. Inspection of these closures to assure initial and long-term integrity of the closure welds presents a challenge to nondestructive testing. The environment is thermally (400 to 10000F) and radioactively (105 R/hr) hot necessitating remote inspection procedures. As a result of research work, ultrasonic test techniques were developed for inspecting the final weld closure of the waste cask. Special transducers, coupling techniques and fixturing were developed and demonstrated in a mockup test facility by remotely examining a 2-in. full penetration weld closure. The examination was performed at room ambient and at a temperature of 2000F. Testing at the desired temperature of 4000F was not completed due to a loss in transducer performance at temperatures in excess of 2000F. Upon completion of the mockup test demonstration, the cask was subjected to a drop test. The ultrasonic results of the pre- and post-examination of two weld closures (the 2-in. full penetration weld and the threaded plug with seal weld) are presented. After the completion of the drop test, both weld closures were radiographed. The radiographs verified the ultrasonic examination and the presence of weld defects in the same areas. Sectioning of the cask closure welds with metallographic verification was not completed at the time of this writing. As a result of the experience gained from the Retrievable Surface Storage Facility (RSSF) storage cask program, recommendations pertaining to the nondestructive engineering development program for Spent Unreprocessed Fuel (SURF) storage casks are presented

  14. CFD analysis of a cask for spent fuel dry storage: Model fundamentals and sensitivity studies

    International Nuclear Information System (INIS)

    Highlights: • A dry storage cask has been evaluated by a CFD code, FLUENT 14. • An alternative methodology for thermal-fluid dynamic modeling has been performed. • Fuel maximum temperature obtained is around 50 K below the regulation limit (673 K). • Even in the most unfavorable heat load distribution temperature increase is smaller than 4%. - Abstract: Dry storage technology must ensure spent fuel cooling under any conditions. This turns thermo-fluid dynamics within dry storage casks a key aspect to investigate, as it would heavily affect fuel rod temperatures. This paper introduces a Computational Fluid Dynamic (CFD) model and analyses of a HI-STORM 100S cask with FLUENT 14.0. Fuel assemblies have been modeled as a porous medium characterized by a thermal conductivity and pressure drop that have been derived from specific approximations, algorithms and methods. This approach has been verified by comparing its results to those published by Holtec International for the HI-STORM cask. The application of the 3D model to HI-STORM 100S cask type under normal conditions, confirms that fuel maximum temperatures more than about 50 K below the regulation limit (673 K) should be expected. In addition, the effect on these results of aspects such as cask design (inlet/outlet orientation), heat load (regionalization) and local climate (external temperature), have been explored. The results indicate that the most relevant factor is heat load distribution and that, even in the most unfavorable regionalization feasible, temperature increase is smaller than 4%. Nonetheless, it should be highlighted that thermal margin to regulatory setting might be reduced down to around 40%

  15. Dry interim spent fuel storage casks. Licensing, evaluation and operational experience

    International Nuclear Information System (INIS)

    The German concept for the external dry interim storage of spent fuel and high level wastes is based on the used of monolithic ductile iron casks which are licensed according to the transport regulations and the national Atomic Energy Act. The casks ensure the safe confinement of the radioactive inventory over long term storage periods of up to 40 years. Essential for that purpose is the double barrier containment system, consisting of two independent lids sealed with long term resistant metallic gaskets and equipped with an interspace pressure monitoring device. Since the establishment of this dry interim storage concept in Germany in the early 1980s, a great deal of experience has been accumulated and now spent fuel elements from the THTR reactor at Hamm-Uentrop and from the AVR research reactor at Juelich are loaded into CASTOR-THTR/AVR casks under dry conditions and stored in the licensed external dry interim storage facilities in Ahaus and Juelich. These are now routine procedures that started in 1992 and has so far comprise more than 200 casks. A great deal of operational experience exists and has also been gained in process optimization without any serious problems. Much more difficult are the drying and evacuation procedures for casks loaded under wet conditions in the spent fuel storage pond of a nuclear power plant. In this case, special operational procedures involving humidity measurements are applied. Different loading operations in several German power plants have been carried out since 1982 and the first wet loaded cask proposed for storage in the licensed external dry interim storage facility at Gorleben came into operation in July 1994. (author). 4 refs, 5 figs, 1 tab

  16. ANALISIS GEN HAEMAGGLUTININ PADA VIRUS CAMPAK LIAR

    Directory of Open Access Journals (Sweden)

    Subangkit Subangkit

    2015-05-01

    Full Text Available AbstrakPenyakit Campak disebabkan oleh virus campak yang termasuk genus Morbilivirus dan Family Paramyxoviridae. Penyakit campak masih menjadi masalah kesehatan karena masih ditemukan Kejadian Luar Biasa (KLB di Indonesia. Salah satu penyebab terjadinya KLB tersebut diduga sebagaiakibat perbedaan antigenesitas antara strain vaksin yang digunakan dengan strain virus campak liar yang beredar di Indonesia. Penelitian ini bertujuan mendapatkan gambaran tentang karakteristik genetik gen Haemagglutinin virus campak liar yang ada di Indonesia. Spesimen yang digunakan sebanyak 27 isolat virus penyebab KLB dari 17 propinsi selama periode tahun 2003-2010. Isolat virus dilakukan pemeriksaan secara RT-PCR dan sekuensing dengan metode Sanger. Hasil sekuensing dianalisis dengan menggunakan perangkat lunak Bioedit 7.0 dan MEGA 4.0. Hasil penelitian didapatkan perbedaan 10 asam amino antara virus campak strain vaksin CAM-70 dan virus campak liar pada posisi D416N; K424T; V451M; N455T; V466I; I473T; F476L; Y481S atau Y481N; H495N; G505D. Kesimpulan penelitian ini adalah terdapat perbedaan karakteristik genetik antara virus campak liar di Indonesia berbeda dengan strain virus vaksin CAM-70.Kata kunci : Campak, Analisis Molekuler, Hemagglutinin, CD46AbstractMeasles is caused by virus belonging to the genus Morbilivirus and Family Paramyxoviridae. Measles is still a public health problem because outbreak of measles still found in Indonesia. Outbreak is suspected as a result of differences in antigenicity between vaccine strains used with wild-type measles virus strains circulating in Indonesia. This study aims to get genetic characteristics of wild-type measles virus haemagglutinin gene in Indonesia. The specimens were used 27 viral isolates from 17 provinces period 2003-2010. Viral isolates examined by RT-PCR and sequencing with Sanger method. Sequencing analysis were conducted using Bioedit 7.0 and MEGA 4.0 software. The results showed 10 amino acid differences

  17. Ageing of a neutron shielding used in transport/storage casks

    Energy Technology Data Exchange (ETDEWEB)

    Nizeyiman, Fidele; Alami, Aatif; Issard, Herve; Bellenger, Veronique [TN International, 1 rue des herons, Montigny le Bretonneux, 78054 Saint Quentin en Yvelines (France); Laboratoire PIMM, Arts and Metiers ParisTech, 151 Bd de l' Hopital, 75013 Paris (France)

    2012-07-11

    In radioactive materials transport/storage casks, a mineral-filled vinylester composite is used for neutron shielding which relies on its hydrogen and boron atoms content. During cask service life, this composite is mainly subjected to three types of ageing: hydrothermal ageing, thermal oxidation and neutron irradiation. The aim of this study is to investigate the effect of hydrothermal ageing on the properties and chemical composition of this polymer composite. At high temperature (120 Degree-Sign C and 140 Degree-Sign C), the main consequence is the strong decrease of mechanical properties induced by the filler/matrix debonding.

  18. TMI-2 (Three-Mile Island-Unit 2) rail cask and railcar maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Tyacke, M.J.; Ayers, A.L., Jr.; Ball, L.J.; Anselmo, A.A.

    1988-02-01

    This paper describes the NuPac 125-B cask system (i.e., cask and railcar), and the maintenance and inspection requirements for that system. The paper discusses the operations being done to satisfy those requirements and how, in some cases, it has been efficient for the operations to be more rigorous than the requirements. Finally, this paper discusses the experiences gained from those operations and how specific hardware and procedural enhancements have resulted in a reliable and continuous shipping campaign. 2 refs., 2 figs.

  19. A structural analysis on the KN-12 spent nuclear fuel transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dew Hey [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Lee, Jae Hyung; Na, Jae Yun [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-08-15

    In this study, safety of the spent nuclear fuel cask KN-12 which is developed in 2000 is evaluated for hypothetical accidents conditions such as free drop, puncture, fire accident and water immersion. Finite element code ABAQUS/Explicit is used to compare with safety analysis report of the GNB in which analysis is performed with LS-DYNA3D for hypothetical accident conditions. Through this study, the safety of KN-12 is evaluated by comprehensive structural analysis. The capability and technological advancement of Korean community on the analysis and structural assessment of the cask will be improved. Also people's anxiety about radioactive dangers will be eliminated.

  20. Safety Analysis of Dual Purpose Metal Cask Subjected to Impulsive Loads due to Aircraft Engine Crash

    Science.gov (United States)

    Shirai, Koji; Namba, Kosuke; Saegusa, Toshiari

    In Japan, the first Interim Storage Facility of spent nuclear fuel away from reactor site is being planned to start its commercial operation around 2010, in use of dual-purpose metal cask in the northern part of Main Japan Island. Business License Examination for safety design approval has started since March, 2007. To demonstrate the more scientific and rational performance of safety regulation activities on each phase for the first license procedure, CREPEI has executed demonstration tests with full scale casks, such as drop tests onto real targets without impact limiters(1) and seismic tests subjected to strong earthquake motions(2). Moreover, it is important to develop the knowledge for the inherent security of metal casks under extreme mechanical-impact conditions, especially for increasing interest since the terrorist attacks from 11th September 2001(3)-(6). This paper presents dynamic mechanical behavior of the metal cask lid closure system caused by direct aircraft engine crash and describes calculated results (especially, leak tightness based on relative dynamic displacements between metallic seals). Firstly, the local penetration damage of the interim storage facility building by a big passenger aircraft engine crash (diameter 2.7m, length 4.3m, weight 4.4ton, impact velocity 90m/s) has been examined. The reduced velocity is calculated by the local damage formula for concrete structure with its thickness of 70cm. The load vs. time function for this reduced velocity (60m/s) is estimated by the impact analysis using Finite Element code LS-DYNA with the full scale engine model onto a hypothetically rigid target. Secondly, as the most critical scenarios for the metal cask, two impact scenarios (horizontal impact hitting the cask and vertical impact onto the lid metallic seal system) are chosen. To consider the geometry of all bolts for two lids, the gasket reaction forces and the inner pressure of the cask cavity, the detailed three dimensional FEM models are

  1. Dry Storage Casks Monitoring by Means of Ultrasonic Tomography

    Science.gov (United States)

    Salchak, Y.; Bulavinov, A.; Pinchuk, R.; Lider, A.; Bolotina, I.; Sednev, D.

    Spent nuclear fuel (SNF) is one of the most hazardous types of nuclear power plant waste. This fact emphasizes the importance of careful handling and storage of SNF. There are two current state-of-the art technologies of SNF storage facility: wet and dry. It is important to mention that IAEA does not determine which kind of handling strategy should be chosen, however it is noted that dry storage of SNF could be used for one hundred years. Mining and Chemical Enterprise (MCE) is one of the leading Russian companies that deals exclusively with the dry storage of SNF. This company has implemented a long-term storage scheme. At the same time MCE faced the challenge of nondestructive monitoring of the degradation process of structural material of cask and its sealing with weld seam. Currently, X-ray testing is used for this purpose but in order to provide an effective nonradioactive method of monitoring MCE has initiated a collaborative R&D project with TPU supported by the Russian Government. Ultrasonic industrial tomography technique was proposed as the solution. The method is based on application of phased and sparse arrays transducer with real-time visualization algorithm. Received acoustic data is processed and realized by means of Sampling Phased Array technology which is a collaborative development of TPU and I-Deal Technology, GmbH. The multichannel ultrasonic set-up of immersion control was assembled for performing testing of seven experimental specimens with representative defects (side drill-holes, notches, natural welding flaws). X-ray tomography of high-resolution was chosen as the reference method. All indications were successfully reconstructed in B and C-scans and 3D image. The next step is to automate the monitoring procedure completely and to introduce an evaluation tool for current flaw state and prediction of its further behavior.

  2. ANALISIS MODEL KEPUASAN TERHADAP PEMBELIAN ULANG

    Directory of Open Access Journals (Sweden)

    Naili Farida

    2014-09-01

    Full Text Available Tujuan penelitian ini yaitu untuk menguji pengaruh brand equity, nilai pelanggan dan lifestyle yang dimediasi kepuasan konsumen terhadap pembelian ulang produk gadget di Kota Semarang. Responden dalam penelitian ini adalah seluruh konsumen yang telah melakukan pembelian produk gadget merek Samsung, sebanyak 120 orang. Teknik sampling yang digunakan yaitu purposive sampling. Teknik analisis yang digunakan yaitu Partial Least Square (PLS. Hasil penelitian menunjukkan bahwa terdapat pengaruh brand equity terhadap kepuasan, namun dalam konteks lifestyle dan nilai pelanggan tidak ada pengaruh terhadap kepuasan. Sedangkan kepuasan menunjukkan adanya pengaruh terhadap pembelian ulang. Diharapkan dari hasil penelitian ini mampu meningkatkan pembelian ulang melalui brand equity, lifestyle dan nilai pelanggan dalam konteks kepuasan. The objective of the research was to test the influence of brand equity, customer value and lifestyle which were mediated by consumer satisfaction toward re-purchasing of gadget products in Semarang. The respondents of the study were all consumers who have purchased Samsung gadget products. It was a purposive sampling study with 120 respondents. The data were analyzed by Partial Least Square (PLS. The result of the study showed that there was an influence of brand equity toward satisfaction, but in the lifestyle context and customer value did not give any influence toward satisfaction. Whereas; the satisfaction gave influence toward re-purchase. It is expected that this study can increase the repurchasethrough brand equity, lifestyle and customer values in the satisfaction context.

  3. Analisis SWOT untuk Penentuan Strategi Optimalisasi Infrastruktur

    Directory of Open Access Journals (Sweden)

    Sri Wahyuningsih

    2012-12-01

    Full Text Available Undang-undang No.38 tahun 2009 tentang Pos memberikan keleluasaan penyelenggara pos untuk melakukan pengembangan produksi serta peningkatan infrastruktur, yang mendukung daya saing masing-masing penyelenggara pos. Sebagai badan usaha milik negara, PT.Pos Indonesia memiliki jaringan terintegrasi sampai ke pedesaan dan daerah transmigrasi, yang diketahui hampir 100% jaringan dibangun oleh pemerintah. Jaringan Pos adalah jaringan fisik maupun virtual untuk mendukung terselenggaranya layanan pos. Jumlah titik layanan mencapai 24 ribu titik layanan yang menjangkau hampir 100% kecamatan dan 42 % kelurahan/desa. Untuk mengetahui strategi dalam rangka optimalisasi infrastruktur di PT.Pos Indonesia, dilakukan penelitian dengan pendekatan kualitatif dan analisis SWOT dan hasilnya penggambaran pada Matrik Grand Strategi, posisi pada  Kuadran I (positif, positif, artinya, PT.Pos Indonesia khususnya untuk KPRK Jakarta Pusat10000, KPRK Jakarta Utara14000 dan KPRK Bandung 40000 manajemen organisasinya sudah solid, dan banyak mempunyai peluang. Ekspansi dapat dilanjutkan untuk memperbesar pertumbuhan dan pengembangan produk, namun aspek SDM masih perlu ditingkatkan.

  4. ANALISIS PERBANDINGAN SPAP, IAS DAN SPKN

    Directory of Open Access Journals (Sweden)

    Maylia Pramono Sari

    2012-03-01

    Full Text Available Dalam mewujudkan Good Goveronance  akan dilakukan analisis perbandingan mengenai standar pemeriksaan nasional yaitu SPAP, ISA, dan SPKN untuk menemukan standar yang lebih tepat dan lebih lengkap untuk pemeriksaan sektor publik. Persamaan SPKN dan ISS  adalah tanggungjawab manajemen dan tanggungjawab auditor sedangkan perbedaannya adalah badan yang menerbitkannya. Perbedaan yang mendasar terletak pada karakter pemeriksaan BPK. Karakter tersebut adalah keharusan pemeriksa BPK untuk merancang prosedur pemeriksaan terhadap kepatuhan yang terkait dengan pemeriksaan yang dilakukan dan waspada atas penyimpangan lainnya. Abstract To realize Good Governance, it needs a comparative analysis for finding out the national standardization of SPAP, ISA, and SPKN . It is used for auditing public sector. SPKN and ISA have similarities and difference. Both SPKN and ISA have the same responsibility of management and audit. SPKN and ISA  have different characteristics in auditing. The compliance of audit procedure is obligatory for  BPK. Moreover, it becomes the strength of BPK RI and the one that make BPK and KAP different.Keywords: SPAP; ISA; SPKN

  5. Analisis Tingkat Kesejahteraan Masyarakat Pesisir Di Kecamatan Medan Labuhan

    OpenAIRE

    Ismail, Fakhri

    2013-01-01

    Tujuan penelitian ini adalah mengetahui tingkat kesejahteraan masyarakat pesisir di Kecamatan Medan Labuhan dengan menggunakan data primer untuk 100 responden yang mewakili seluruh populasi masyarakat pesisir di Kecamatan Medan Labuhan. Pengumpulan data dilakukan dengan menggunakan daftar kuesioner. Metode analisis yang digunakan adalah deskriptif kualitatif. Data yang terkumpul diolah dan disajikan dalam bentuk tabel Hasil penelitian menunjukkan bahwa masyarakat pesisir di Kecamatan Meda...

  6. Un’analisi strutturale: Il corvo bianco di P. Bigongiari

    Directory of Open Access Journals (Sweden)

    Luciano Vitacolonna

    2011-12-01

    Full Text Available Presentiamo qui un'analisi strutturale della lirica Il corvo bianco di Piero Bigongiari. Vengono così messi in luce sia gli strati del testo, sia le relazioni intra- e intertestuali. Attraverso lo scandaglio degli aspetti compositivi della poesia ne proponiamo infine una complessiva lettura simbolica.

  7. Analisis Desain Layar 3D Menggunakan Pengujian Pada Wind Tunnel

    Directory of Open Access Journals (Sweden)

    Danang Priambada

    2012-09-01

    Full Text Available Jurnal ini merupakan hasil analisis design layar menggunakan konsep dasar NACA 0012 dengan memprediksi perilaku dinamika fluida pada model layar rigid. Sarana yang dapat digunakan untuk memprediksi perilaku fluida udara yaitu menggunakan pengujian model pada wind tunnel. Tujuan penelitian ini adalah untuk mengetahui bagaimana pengaruh bentuk layar dengan faktor variasi desain berupa perbedaan penampang samping layar dengan aspek ratio yang sama serta pengaruh sudut serang terhadap gaya dorong yang dihasilkan. Analisis data dilakukan melalui pengujian wind tunnel subsonic. Pengujian dilakukan pada sudut serang 10ο, 15ο, 20ο. Dari pengujian model layar pada wind tunnel subsonic akan dihasilkan data-data yang berpengaruh pada pada layar seperti lift, drag, dan gaya resultan. Sehingga dapat dilakukan analisis terhadap model layar untuk mendapatkan desain layar yang memiliki gaya dorong yang paling optimal. Berdasarkan analisis dari wind tunnel maka didapatkan titik Stall pada sudut serang 15ο. Nilai koefisien lift terbesar terjadi pada model 3 sudut serang 15ο. Efisiensi layar terbesar juga terjadi pada model 3 yaitu sebesar 5,149.

  8. Analisis Investasi pada Proyek Pembangunan Apartemen Bale Hinggil Surabaya

    Directory of Open Access Journals (Sweden)

    Ofianto Ofianto Wahyudhi

    2014-03-01

    Full Text Available Pertumbuhan penduduk Surabaya dari tahun ke tahun terus mengalami peningkatan. Hal ini berpengaruh terhadap meningkatnya kebutuhan tempat tinggal yang layak. Maka dari itu salah satu developer PT Tlatah Gema Nugraha ingin membangun sebuah  hunian vertikal yang bernama Apartemen Bale Hinggil. Beberapa analisis sangat diperlukan melihat dari peningkatan pembangunan apartemen di Surabaya yang begitu pesat, membuat tingkat persaingan menjadi tinggi. Selain itu kecenderungan okupansi apartemen yang terus-menerus  menurun sekarang ini. Tugas akhir ini bertujuan  untuk  mendapatkan alternatif pendapatan  yang optimal dari pemilihan alternatif pendapatan antara apartemen sewa, apartemen jual, dan gabungan keduanya. Untuk penilaian alternatif sistem pendapatan digunakan analisis aruskan dengan penilaian kelayakan investasi menggunakan metode NPW dan IRR. Pada analisis kelayakan investasi didapatkan nilai NPV dan IRR tertinggi pada sistem jual yaitu sebesar Rp155.907.406.750,-  untuk NPV dan 46,62% untuk IRR. Untuk analisis sensitivitas menunjukkan alternatif pendapatan pada sistem jual merupakan alternatif yang memiliki rentan lebih aman terhadap kelayakannya jika terjadi perubahan variabel investasi. Sedangkan untuk alternatif pendapatan pada sistem gabungan memiliki tingkat sensitivitas tinggi sehingga jika terjadi perubahan variabel investasi dapat membuat alternatif pendapatan dengan sistem gabungan menjadi tidak layak.

  9. Interactions between cask components and content of packaging for the transport of radioactive material during drop tests

    International Nuclear Information System (INIS)

    This paper describes the analytical, numerical and experimental investigations on the phenomenon of interactions between cask components and content of packages for the transport of radioactive material during drop tests required according to the IAEA Regulations for the Safe Transport of Radioactive Material. Radial and axial gaps between cask components and content are usually necessary for thermal reasons but larger gaps can exist because of the geometrical dimensions of the specified content. Consequently interactions between content and cask components (lid system, cask body, etc.) are possible and can not be excluded during drop tests. Interactions in this context are relative movements between cask and content which are mainly due to elastic spring effects after releasing the cask for the free drop. These relative movements can cause interior collisions between content and cask during the main impact of the package onto the unyielding target. Drop tests with various types of Type A and Type B packages fully instrumented with strain gauges and accelerometers showed that these interactions respectively interior collisions can be considerable relating to high forces acting on cask lids, lid bolts and the content. Of course the real quantitative consequences of the interactions depend upon different conditions, among others the drop orientation, the design characteristics of the impact limiters, the dimensions of the gaps, the material characteristics of the contents, etc. . In order to investigate more precisely the phenomenon of interactions BAM carried out finite element calculations for the named casks using the ABAQUS/ Standard and ABAQUS/ Explicit computer code comparing them with results obtained from experiments. Additionally, tests with a simplified model instrumented with accelerometers were carried out accompanied by finite element calculations and analytical calculations using MATHEMATICA. The investigations on the mentioned phenomena of interaction

  10. Study on the key technologies of the Transfer Equipment Cask for Tokamak Equator Port Plug

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Buyun, E-mail: ayun@iim.ac.cn [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Gao, Lifu [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Cao, Huibin; Sun, Jian [Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Sun, Yuxiang; Song, Quanjun; Ma, Chengxue; Chang, Li; Shuang, Feng [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2014-12-15

    Highlights: • Design on Intelligent Air Transfer System (IATS) for Transfer Equipment Cask (TECA). • A rhombic-like parallel robot for docking with minimum misalignment. • Design on electro-hydraulic servo system of the TECA for Tokamak Equator Port Plug (TEPP) manipulation. • A control architecture with several algorithms and information acquired from sensors could be used by the TECA for Remote Handling (RH). - Abstract: The Transfer Equipment Cask (TECA) is a key solution for Remote Handling (RH) in Tokamak Equator Port Plug (TEPP) operations. From the perspectives of both engineering and technical designs of effective experiments on the TEPP, key technologies on these topics covering the TECA are required. According to conditions in ITER (International Thermonuclear Experimental Reactor) and features of the TEPP, this paper introduces the design of an Intelligent Air Transfer System (IATS) with an adaptive attitude and high precision positioning that transports a cask system of more than 30 tons from the Tokamak Building (TB) to the Hot Cell Building (HCB). Additionally, different actuators are discussed, and the hydraulic power drive is eventually selected and designed. A rhombic-like parallel robot is capable of being used for docking with minimum misalignment. Practical mechanisms of the cask system are presented for hostile environments. A control architecture with several algorithms and information acquired from sensors could be used by the TECA. These designs yield realistic and extended applications for the RH of ITER.

  11. Study on the key technologies of the Transfer Equipment Cask for Tokamak Equator Port Plug

    International Nuclear Information System (INIS)

    Highlights: • Design on Intelligent Air Transfer System (IATS) for Transfer Equipment Cask (TECA). • A rhombic-like parallel robot for docking with minimum misalignment. • Design on electro-hydraulic servo system of the TECA for Tokamak Equator Port Plug (TEPP) manipulation. • A control architecture with several algorithms and information acquired from sensors could be used by the TECA for Remote Handling (RH). - Abstract: The Transfer Equipment Cask (TECA) is a key solution for Remote Handling (RH) in Tokamak Equator Port Plug (TEPP) operations. From the perspectives of both engineering and technical designs of effective experiments on the TEPP, key technologies on these topics covering the TECA are required. According to conditions in ITER (International Thermonuclear Experimental Reactor) and features of the TEPP, this paper introduces the design of an Intelligent Air Transfer System (IATS) with an adaptive attitude and high precision positioning that transports a cask system of more than 30 tons from the Tokamak Building (TB) to the Hot Cell Building (HCB). Additionally, different actuators are discussed, and the hydraulic power drive is eventually selected and designed. A rhombic-like parallel robot is capable of being used for docking with minimum misalignment. Practical mechanisms of the cask system are presented for hostile environments. A control architecture with several algorithms and information acquired from sensors could be used by the TECA. These designs yield realistic and extended applications for the RH of ITER

  12. Development of strain gauge evaluation channels for use in dynamic testing of shipping casks

    International Nuclear Information System (INIS)

    The Transportation System Development Department at Sandia National Laboratories (SNL) frequently evaluates the structural response of casks being developed to transport radioactive materials. A major part of this activity includes gathering instrumentation data from dynamic impact tests of cask models. The acquisition of reliable, high-quality instrumentation data is an important component of cask certification. One method to evaluate instrumentation error during testing is to include evaluation channels for the various structural transducers. Evaluation channels have been produced by some manufacturers of accelerometers used for structural evaluations of casks and are commercially available. These particular devices produce very low output or no output to applied shock acceleration. However, it was found that a packaged strain gauge evaluation channel is not commercially available. Consequently, strain gauge evaluation channels have been developed at SNL to evaluate non-strain-induced resistance changes from environmental factors that could affect resistance strain measurement data. These unwanted nonstrain-induced resistance changes could be caused, for example, by resistance changes in the interconnecting cabling, electromagnetic noise, or grounding effects

  13. Imaging Spent Fuel in Dry Storage Casks with Cosmic Ray Muons

    Energy Technology Data Exchange (ETDEWEB)

    Durham, J. Matthew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dougan, Arden [National Nuclear Security Administration (NNSA), Washington, DC (United States)

    2015-11-05

    Highly energetic cosmic ray muons are a natural source of ionizing radiation that can be used to make tomographic images of the interior of dense objects. Muons are capable of penetrating large amounts of shielding that defeats typical radiographic probes like neutrons or photons. This is the only technique which can examine spent nuclear fuel rods sealed inside dry casks.

  14. 75 FR 49813 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1, Confirmation of...

    Science.gov (United States)

    2010-08-16

    ... 3150-AI86 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1, Confirmation of..., 2010, for the direct final rule that was published in the Federal Register on June 15, 2010 (75 FR 33678). This direct final rule amended the NRC's spent fuel storage regulations at 10 CFR 72.214...

  15. 77 FR 9515 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-02-17

    ... Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a... RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear... Commission) is amending its spent fuel storage regulations by revising the Holtec International HI-STORM...

  16. 75 FR 42292 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

    Science.gov (United States)

    2010-07-21

    ... for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also... COMMISSION 10 CFR Part 72 RIN 3150-AI88 List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6... Commission (NRC) is amending its spent fuel storage regulations by revising the NAC International Inc....

  17. 76 FR 9381 - Notice of Availability of Interim Staff Guidance Documents for Spent Fuel Storage Casks

    Science.gov (United States)

    2011-02-17

    ... COMMISSION Notice of Availability of Interim Staff Guidance Documents for Spent Fuel Storage Casks AGENCY... Gordon, Structural Mechanics and Materials Branch, Division of Spent Fuel Storage and Transportation... performing technical reviews of spent fuel storage and transportation packaging licensing actions.'' This...

  18. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false List of approved spent fuel storage casks. 72.214 Section... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.214 List of approved...

  19. Impact stress reduction by shell splitting in cask for transporting radioactive material

    International Nuclear Information System (INIS)

    Highlights: • High impact stress in shell of a container for transporting radioactive material. • Reduction of impact stress by splitting shell into multiple parts. • Impact simulations on simple objects to prove benefits of shell splitting. • Explanation based on theory of bending of simply supported beam. • Impact simulations on a simple cask showing up to 21% reduction in maximum stress. - Abstract: Casks designed for transporting radioactive material are mandated to withstand drop from specific heights on hard ground. The maximum internal stress in the shell of the cask after such an impact needs to be as low as possible to ensure safety of the material being transported. This paper investigates the concept of splitting the shell of the radioactive transport container into multiple layers to reduce these stresses after impact. Different geometrical configurations which are likely to be encountered while designing such containers have been studied through plane 2D and 3D finite element analysis and the efficacy of this idea has been explored on each of them. Considerable reduction of stress has been reported and an explanation based on elastic deformation of layered beams has been suggested. Simulations on a cask with the currently prevalent design also show the benefit of implementing this idea

  20. A Stylistic Analysis on Edgar Allan Poe's The Cask of Amontillado

    Institute of Scientific and Technical Information of China (English)

    杨赛菲

    2016-01-01

    The Cask of Amontillado is one of Poe's best-known horror short stories. Based on Stylistics, this paper attempts to analyze this story from the aspects of themes, characterization, point of view, syntactic and lexical features, to reveal Poe's excellent skills and the artistic charm.

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

    International Nuclear Information System (INIS)

    A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. To overcome modeling difficulties arising from the complexity of geometry in large PWR metal casks, a multiple cylinder model is used to calculate the temperature profile of a cylindrical cask body in the first step analysis. In the second step analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three-dimensional conduction analysis model. An existing HEATING 7.2f code has been used in the present two-step numerical analyses. Effects of aluminium heat transfer fin and the cask ambient conditions on the maximum fuel temperature have been examined as a parametric study. A comparison between the predicted maximum fuel temperature and the data of Nuclear Assurance Corporation Storage and Transportation Canister Safety Analysis Report (NAC-STC SAR) shows good agreement

  2. Validation of elastic-plastic computer analyses for use in nuclear waste shipping cask design

    International Nuclear Information System (INIS)

    GA Technologies designed the Defense High Level Waste (DHLW) Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Laboratories (SNL). The DHLW cask has a thick-walled stainless steel body and incorporates integral stainless steel impact limiters that protect the two ends of the cask during the hypothetical accident condition 30-ft free drop. These integral impact limiters absorb the drop energy through gross plastic deformations. GA used elastic-plastic computer codes developed at Los Alamos and Lawrence Livermore Laboratories, HONDOII and DYNA3D, to analyze for this non-linear behavior. In order to evaluate the analyses, GA developed elastic-plastic stress criteria that were adapted from the ASME Boiler and Pressure Vessel Code, Division I, Section III. This innovative design and analytical approach required test verification. Therefore, SNL performed 30-ft drop and puncture tests on a half-scale model of the DHLW cask. The testing confirmed that the analytical approach works and results in a safe, conservative design

  3. Studies and research concerning BNFP: operational assessment of the FSV-1 (HTGR) spent fuel shipping cask in alternate modes

    International Nuclear Information System (INIS)

    This report presents an operational assessment of the FSV-1 spent fuel shipping cask which was developed by General Atomic specifically for the Fort St. Vrain reactor. Of primary interest is the adaptation to underwater loading and unloading of light-water-reactor (LWR) fuel in this cask which was designed for the dry-handling of high temperature gas cooled reactor (HTGR) fuel. Also presented is a concept for a system to compare the pros and cons of wet and dry handling of this and other casks

  4. Structural dimensioning of dual purpose cask prototype; Dimensionamento estrutural de prototipo de casco de duplo proposito

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Luiz Leite da; Mourao, Rogerio Pimenta; Lopes, Claudio Cunha [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)]. E-mail: silvall@cdtn.br; mouraor@cdtn.br; ccl@cdtn.br

    2005-07-01

    The structural dimensioning of a Type B(U) dual purpose cask prototype is part of the scope of work of the Brazilian institute CDTN in the IAEA regional project involving Latin American countries which operate research reactors (Argentina, Brazil, Chile, Mexico and Peru). In order to meet the dimensional and operational characteristics of the reactor facilities in these countries, a maximum weight of 10.000 kgf and a maximum dimension of 1 m in at least one direction were set for the cask. With these design restrictions, the cask's payload is either 21 MTR or 78 TRIGA fuel elements. The cask's most important components are main body, primary and secondary lids, basket and impact limiters. The main body has a sandwich-like wall with internal and external layers made of AISI 304 stainless steel with lead in-between. The lead provides biological shielding. The primary lid is similarly layered, but in the axial direction. It is provided with a double system of metallic rings and has ports for pressurization, sampling and containment verification. The secondary lid has the main function of protecting the primary lid against mechanical impacts. The basket structure is basically a tube array reinforced by bottom plate, feet and spacers. Square tubes are used for MTR elements and circular tubes for TRIGA elements. Finally, the impact limiters are structures made of an external stainless steel thin covering and a filling made of the wood composite OSB - Oriented Strand Board. The prototype is provided with bottom and top impact limiters, which are attached to each other by means of four threaded rods. The limiters are not rigidly attached to the cask body. A half scale cask model was designed to be submitted to a testing program. As its volume scales down to 1:8, the model weight is 1,250 kgf. This paper presents the methodology for the preliminary structural dimensioning of the critical parameters of the cask prototype. Both normal conditions of operation and

  5. (Validation of) computational fluid dynamics modeling approach to evaluate VSC-17 dry storage cask thermal designs

    International Nuclear Information System (INIS)

    This paper presents results from a numerical analysis of the thermal evaluation of a Ventilated Concrete Storage Cask VSC-17 system. Three-dimensional simulations are performed for the VSC-17 system, and the results are compared to experimental data. The VSC-17 is a concrete-shielded spent nuclear fuel (SNF) cask system designed to contain 17 pressurized water reactor (PWR) fuel assemblies for storage and transportation. The system consists of a ventilated concrete cask (VCC) and a multi-assembly sealed basket (MSB). The VCC is a concrete cylindrical vessel, fabricated as a single piece and fitted with a flat plate at the bottom. The concrete cask provides structural support, shielding, and natural convection cooling for the MSB. The MSB has an outer steel shell and an inner fuel guide sleeve assembly that holds canisters containing spent fuel rods. Cooling airflow inside the concrete cask is driven by natural convection. Heat transfer in the cask is a complicated process because of the inherent complexity of the geometry and the fixed and natural convection induced by the radioactive decay process. Other factors that contribute to the overall heat transfer include the heat generation by the spent fuel, the thermal boundary condition, the filling medium within the MSB, and the vertical or horizontal orientation of the cask. Proper thermal analysis of dry storage casks is important for accurate estimation of the peak fuel temperature and peak cladding temperature (PCT). Proper estimation of PCT ensures the integrity of cladding and is important for safety evaluation of independent spent fuel storage installations. Accurate estimation of the peak fuel temperature and peak cladding temperature ensures the integrity of the cladding. The spent nuclear fuel may be exposed to air and oxidize if the cladding is damaged and thus increase the potential for release of radioactivity. In the current analysis, numerical simulations are carried out using the computational fluid

  6. Thermo-mechanical finite element analyses of bolted cask lid structures

    Energy Technology Data Exchange (ETDEWEB)

    Wieser, G.; Qiao Linan; Eberle, A.; Voelzke, H. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

    2004-07-01

    The analysis of complex bolted cask lid structures under mechanical or thermal accident conditions is important for the evaluation of cask integrity and leak-tightness in package design assessment according to the Transport Regulations or in aircraft crash scenarios. In this context BAM is developing methods based on Finite Elements to calculate the effects of mechanical impacts onto the bolted lid structures as well as effects caused by severe fire scenarios. I n case of fire it might be not enough to perform only a thermal heat transfer analysis. The complex cask design in connection with a severe hypothetical time-temperature-curve representing an accident fire scenario will create a strong transient heating up of the cask body and its lid system. This causes relative displacements between the seals and its counterparts that can be analyzed by a so-called thermo-mechanical calculation. Although it is currently not possible to correlate leakage rates with results from deformation analyses directly an appropriate Finite Element model of the considered type of metallic lid seal has been developed. For the present it is possible to estimate the behaviour of the seal based on the calculated relative displacements at its seating and the behaviour of the lid bolts under the impact load or the temperature field respectively. Except of the lid bolts the geometry of the cask and the mechanical loading is axial-symmetric which simplifies the analysis considerably and a two-dimensional Finite Element model with substitute lid bolts may be used. The substitute bolts are modelled as one-dimensional truss or beam elements. An advanced two-dimensional bolt submodel represents the bolts with plane stress continuum elements. This paper discusses the influence of different bolt modelling on the relative displacements at the seating of the seals. Besides this, the influence of bolt modelling, thermal properties and detail in geometry of the two-dimensional Finite Element models on

  7. High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-02-27

    EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

  8. PRELIMINARY REPORT: EFFECTS OF IRRADIATION AND THERMAL EXPOSURE ON ELASTOMERIC SEALS FOR CASK TRANSPORTATION AND STORAGE

    Energy Technology Data Exchange (ETDEWEB)

    Verst, C.; Skidmore, E.; Daugherty, W.

    2014-05-30

    A testing and analysis approach to predict the sealing behavior of elastomeric seal materials in dry storage casks and evaluate their ability to maintain a seal under thermal and radiation exposure conditions of extended storage and beyond was developed, and initial tests have been conducted. The initial tests evaluate the aging response of EPDM elastomer O-ring seals. The thermal and radiation exposure conditions of the CASTOR® V/21 casks were selected for testing as this cask design is of interest due to its widespread use, and close proximity of the seals to the fuel compared to other cask designs leading to a relatively high temperature and dose under storage conditions. A novel test fixture was developed to enable compression stress relaxation measurements for the seal material at the thermal and radiation exposure conditions. A loss of compression stress of 90% is suggested as the threshold at which sealing ability of an elastomeric seal would be lost. Previous studies have shown this value to be conservative to actual leakage failure for most aging conditions. These initial results indicate that the seal would be expected to retain sealing ability throughout extended storage at the cask design conditions, though longer exposure times are needed to validate this assumption. The high constant dose rate used in the testing is not prototypic of the decreasingly low dose rate that would occur under extended storage. The primary degradation mechanism of oxidation of polymeric compounds is highly dependent on temperature and time of exposure, and with radiation expected to exacerbate the oxidation.

  9. Behavior of spent fuel and cask components after extended periods of dry storage

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (NRC) promulgated 10 CFR Part 72, Title 10, for the independent storage of spent nuclear fuel and high-level radioactive waste outside reactor spent fuel pools. Part 72 currently limits the license term for an independent spent fuel storage installation to 20 years from the date of issuance. Licenses may be renewed by the Commission at or before the expiration of the license term. Applications for renewal of a license should be filed at least two years prior to the expiration of the existing license. In preparation for possible license renewal, the NRC Office of Nuclear Material and Safeguards, Spent Fuel Project Office, is developing the technical basis for renewals of licenses and Certificates of Compliance for dry storage systems for spent nuclear fuel and high-level radioactive waste at independent spent fuel storage installation sites. An analysis of past performance of selected components of these systems is required as part of that technical basis. In the years 1980 through the early 1990, the Department of Energy (DOE) procured four prototype dry storage casks for testing at the Idaho National Engineering and Environmental Laboratory (INEEL): Castor-V/21, MC-10, TN-24P, and VSC-17. The primary purpose of the testing was to benchmark thermal and radiological codes and to determine the thermal and radiological characteristics of the casks. A series of examinations in 1999 and early 2000 to investigate the integrity of the Castor V/21 cask were undertaken. There is no evidence of significant degradation of the Castor V/21 cask systems important to safety from the time of initial loading of the cask in 1985 up to the time of testing in 1999. (author)

  10. Impact Analyses and Tests of Metal Cask Considering Aircraft Engine Crash - 12308

    International Nuclear Information System (INIS)

    The structural integrity of a dual purpose metal cask currently under development by the Korea Radioactive Waste Management Cooperation (KRMC) is evaluated through analyses and tests under a high-speed missile impact considering the targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from the literature. The missile impact velocity was set at 150 m/s, and two impact orientations were considered. A simplified missile simulating a commercial aircraft engine is designed from an impact load history curve provided in the literature. In the analyses, the focus is on the evaluation of the containment boundary integrity of the metal cask. The analyses results are compared with the results of tests using a 1/3 scale model. The results show very good agreements, and the procedure and methodology adopted in the structural analyses are validated. While the integrity of the cask is maintained in one evaluation where the missile impacts the top side of the free standing cask, the containment boundary is breached in another case in which the missile impacts the center of the cask lid in a perpendicular orientation. A safety assessment using a numerical simulation of an aircraft engine crash into spent nuclear fuel storage systems is performed. A commercially available explicit finite element code is utilized for the dynamic simulation, and the strain rate effect is included in the modeling of the materials used in the target system and missile. The simulation results show very good agreement with the test results. It is noted that this is the first test considering an aircraft crash in Korea. (authors)

  11. ANALISIS FAKTOR DAYA SAING DI KABUPATEN SEMARANG

    Directory of Open Access Journals (Sweden)

    Eka Handriani

    2010-03-01

    Full Text Available Jika perusahaan dapat mengoptimalkan penggunaan sumberdaya maka perusahaan mampu menghasilkan sustained competitive advantage. Tujuan dari penelitian ini adalah untuk membuktikan bahwa faktor internal dan eksternal, keterampilan pengusaha dan strategi yang diterapkan dalam UKM Kabupaten Semarang berpengaruh terhadap keunggulan kompetitif baik secara simultan dan parsial. Pengusaha mikro memiliki potensi besar dalam ekonomi nasional dan lokal sesuai dengan periode pasar bebas. Namun, etos dan persaingan masih dalam tingkat rendah. Penyebab utama dari masalah ini berasal dari faktor tenaga lingkungan, di antaranya adalah faktor internal dan eksternal keterampilan pengusaha, strategi, dan etos. Sebuah strategi yang akurat diperlukan untuk memecahkan masalah ini. Penelitian ini dilakukan pada 60 pengusaha mikro melalui empat variabel dependen dan satu variabel independen dengan menggunakan alat analisis statistik program SPSS. Hasilnya menunjukkan bahwa faktor internal dan eksternal, keterampilan pengusaha, strategi dan etos telah signifikan berpengaruh pada keunggulan kompetitif yang diterapkan oleh pengusaha mikro di Kabupaten Semarang secara parsial maupun simultan. The purpose of this research is to attest that internal and external factor, entrepreneur skill and strategy which applied in the UKM of Kabupaten Semarang are influence to the competitive advantage simultaneously and partially. Micro entrepreneur has great potency in the national and local economic according to this free market period. However, it is ethos and competition still in the low level. The main cause of this problem come up from environmental exertion factor, among them are internal and external factor the entrepreneur skill, strategy, and ethos. An accurate strategy is needed to solve this problem. This research has been done on 60 micro entrepreneurs through four dependent variables and one independent variable using to SPSS statistic program. The result shown that

  12. Feasibility Study For Use Of Commercial Cask Vendor Dry Transfer Systems To Unload Used Fuel Assemblies In L-Area

    Energy Technology Data Exchange (ETDEWEB)

    Krementz, Dan; Rose, David; Dunsmuir, Mike

    2014-02-06

    The purpose of this study is to determine whether a commercial dry transfer system (DTS) could be used for loading or unloading used nuclear fuel (UNF) in L-Basin and to determine if a DTS pool adapter could be made for L-Basin Transfer Pit #2 that could accommodate a variety of DTS casks and fuel baskets or canisters up to 24” diameter.[1, 2] This study outlines the technical feasibility of accommodating different vendor dry transfer systems in the L-Basin Transfer Bay with a general work scope. It identifies equipment needing development, facility modifications, and describes the needed analyses and calculations. After reviewing the L-Basin Transfer Bay area layout and information on the only DTS system currently in use for the Nuclear Assurance Corporation Legal Weight Truck cask (NAC LWT), the authors conclude that use of a dry transfer cask is feasible. AREVA was contacted and acknowledged that they currently do not have a design for a dry transfer cask for their new Transnuclear Long Cask (TN-LC) cask. Nonetheless, this study accounted for a potential future DTS from AREVA to handle fuel baskets up to 18” in diameter. Due to the layout of the Transfer Bay, it was determined that a DTS cask pool adapter designed specifically for spanning Pit #2 and placed just north of the 70 Ton Cask lid lifting superstructure would be needed. The proposed pool adapter could be used to transition a fuel basket up to 24” in diameter and ~11 feet long from a dry transfer cask to the basin. The 18” and 24” applications of the pool adapter are pending vendor development of dry transfer casks that accommodate these diameters. Once a fuel basket has been lowered into Pit #2 through a pool adapter, a basket cart could be used to move the basket out from under the pool adapter for access by the 5 Ton Crane. The cost to install a dry transfer cask handling system in L-Area capable of handling multiple vendor provided transport and dry transfer casks and baskets with

  13. Analysis technology in the thick plate free drop impact, heat and thermal stress of the cask for radioactive material transport

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dew Hey [Korea Institute of Nuclear and Safety, Taejon (Korea, Republic of); Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Choi, Kyung Joo; Choi, Young Jin; Lee, Jae Hyung; Na, Jae Yun; Kim, Seong Jong [Chungnam National Univ., Taejon (Korea, Republic of)

    2002-03-15

    In this study, The regulatory condition and analysis condition is analyzed for thick plate free drop, heat and thermal stress analysis to develop the safety assessment technology. Analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. ANSYS, LS-DYNA3D and ABAQUS is suitable for thick plate free drop, heat and thermal stress analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS for thick plate free drop and the results of ANSYS, LS-DYNA3D and ABAQUS for heat and thermal stress analysis is completely corresponded. And the integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology.

  14. Safety assessment technology on the free drop impact and puncture analysis of the cask for radioactive material transport

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dew Hey [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Lee, Ho Chul; Hong, Song Jin; Choi, Young Jin; Lee, Jae Hyung; Na, Jae Yun [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-03-15

    In this study, the regulatory condition and analysis condition is analyzed for the free drop and puncture impact analysis to develop the safety assessment technology. Impact analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. LS-DYNA3D and ABAQUS is suitable for the free drop and the puncture impact analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS is completely corresponded. And The integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology.

  15. Operations of the LR56 radioactive liquid cask transport system at U.S. Department of Energy sites

    Energy Technology Data Exchange (ETDEWEB)

    Davidson, J.S. [Lockheed Martin Energy Systems, Oak Ridge, TN (United States); Hornstra, D.J. [Performance Development Corp., Oak Ridge, TN (United States); Sazawal, V.K. [NUMATEC, Inc., Bethesda, MD (United States); Clement, G. [SGN, St. Quentin en Yvelines (France)

    1996-06-01

    The LR56 cask system is licensed for use in France under Certificate of Compliance F/309/B(U)F for transport of 4,000-liter volumes of radioactive liquids. Three LR56 cask systems (with modifications for use at Department of Energy (DOE) sites) have been purchased for delivery at the Hanford Site, Oak Ridge National Laboratory (ORNL), and Savannah River Site (SRS). The LR56 cask systems will be used for on-site transfers of Type B quantities of radioactive liquid waste. The ORNL unit will also be used as a Type A packaging for transfers of radioactive liquids between DOE sites. This paper discusses LR56 operating features and the use of the cask system at the three DOE sites.

  16. Safety assessment technology on the free drop impact and puncture analysis of the cask for radioactive material transport

    International Nuclear Information System (INIS)

    In this study, the regulatory condition and analysis condition is analyzed for the free drop and puncture impact analysis to develop the safety assessment technology. Impact analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. LS-DYNA3D and ABAQUS is suitable for the free drop and the puncture impact analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS is completely corresponded. And The integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology

  17. Analysis technology in the thick plate free drop impact, heat and thermal stress of the cask for radioactive material transport

    International Nuclear Information System (INIS)

    In this study, The regulatory condition and analysis condition is analyzed for thick plate free drop, heat and thermal stress analysis to develop the safety assessment technology. Analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. ANSYS, LS-DYNA3D and ABAQUS is suitable for thick plate free drop, heat and thermal stress analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS for thick plate free drop and the results of ANSYS, LS-DYNA3D and ABAQUS for heat and thermal stress analysis is completely corresponded. And the integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology

  18. The state of the Primary Degradation Factors and Models of Concrete Cask in Spent Fuel Dry Storage System

    International Nuclear Information System (INIS)

    In South Korea, a total of twenty nuclear reactors are in operation; the cumulative amount of spent fuel is estimated to be 10,490 MTU in 2009. The full capacity of the waste storage is expected to be saturated in around 2016. However, a national strategy for spent fuel management has not yet been set down and high level waste (HLW) such as spent fuel will have to be stored at-reactor (AR) by re-racking. Recently an worldwide interest on the dry storage has increased especially around U.S. With a perspective of the material of the spent fuel dry storage cask, the system can be divided into two types of metal and concrete casks. The concrete type cask is a very attractive option because of the cost competitiveness of concrete material and its relatively long-term durability. Although the type of metal cask is chosen, the use of cementitious material is inevitable at least for the cask foundation and the facilities for the protection of dry storage structures. Upon being placed, the performance of concrete begins to deteriorate from the intrinsic change of cement and the physical/ chemical environmental conditions. Thus it is necessary to evaluate the durability of a concrete for the increase of reliability and safety of the whole system during the designed life time. Considering the dry storage system of spent fuel is the item which can create a lot of added value, the development of a dry storage cask is usually initiated by private enterprises among developed countries. The detail research results and specific design criteria for the safety assessment of a concrete cask have not been revealed to the public well. In this paper, the major expected degradation factors and related degradation models of concrete casks were investigated as part of the safety assessment by taking account of the site where Korea industrial nuclear power plants are located

  19. Spent-fuel shipping and cask-handling studies in wet and dry environments. Studies and research concerning BNFP

    International Nuclear Information System (INIS)

    A demonstration cask system has been constructed specifically to be used in examining unconventional techniques in handling spent fuel and fuel-hauling casks. This report demonstrates, through a series of photographs, some of these techniques and discusses others. It includes wet and dry operations, loading and unloading horizontally and vertically, mobile on-site carriers that can eliminate the need for some cranes and, in general, many of the operational options that are open in the design of future fuel handling systems

  20. ANALISIS TINGKAT KAPABILITAS SISTEM INFORMASI RUMAH SAKIT BERDASARKAN COBIT 5 (MEA01) PADA RSUD TUGUREJO SEMARANG

    OpenAIRE

    Ariel Bagus Nugroho; Amiq Fahmi

    2015-01-01

    RSUD Tugurejo Semarang merupakan Rumah Sakit kelas B milik pemerintah Provinsi Jawa Tengah yang menyediakan pelayanan jasa untuk masyarakat. Dalam manajemen pengolahan data, RSUD Tugurejo Semarang telah menggunakan sistem informasi rumah sakit. Untuk mengetahui apakah sistem informasi telah berjalan seperti yang diharapkan, maka perlu dilakukan analisis tatakelola teknologi informasi. Pada penelitian ini analisis tatakelola teknologi informasi berfokus pada proses pengawasan, evaluasi dan pen...

  1. TGC36 a dual purpose cask for the transport and interim storage of compacted waste (CSD-C)

    International Nuclear Information System (INIS)

    According to contractual and international obligations, the German Utilities have to return the residues resulting from the reprocessing of nuclear fuel assemblies (compacted hulls and ends) to Germany. The new dual purpose cask TGC36 is a joint product from the two leading companies in the field development and manufactory of nuclear casks in Europe, GNS and TN International, is intended for the transport to the interim storage facility Ahaus and to be stored there for up several years. For the development and the delivery of the TGC36 cask, GNS and TN International formed the AGC Consortium based on German law to combine the special know how of both partners in the most efficient way. The design and the licensing strategy of the TGC36 are introduced in this paper. In conclusions: GNS and TNI have formed a consortium named AGC to design, license and manufacture an innovative cask for the transport and the interim storage of the compacted wastes resulting from the reprocessing of the German spent fuel. This cask has been optimized in order to offer a high capacity of loading, and allows a payload of 36 canisters, leading to a total mass of approximately 116 Mg in transport configuration. The success of this project requires a special effort from both partner companies, members of the consortium, and implies also an efficient management of simultaneous tasks during the licensing period and the manufacturing time of the first items of the cask. (authors)

  2. A comparison of spent-fuel shipping cask response to 10 CFR 71 normal conditions and realistic hot day extremes

    International Nuclear Information System (INIS)

    The structural properties of spent nuclear fuel shipping containers vary as a function of the cask wall temperature. An analysis is performed to determine the effect of a realistic, though bounding, hot day environment on the thermal behavior of spent fuel shipping casks. These results are compared to those which develop under a steady-state application of the prescribed normal thermal conditions of 10CFR71. The completed analysis revealed that the majority of wall temperatures, for a wide variety of spent fuel shipping cask configurations, fall well below those predicted by using the steady-state application of the regulatory boundary conditions. It was found that maximum temperatures at the cask surface occasionally lie above temperatures predicted under the regulatory condition. This is due to the conservative assumptions present in the ambient conditions used. The analysis demonstrates that diurnal temperature variations which penetrate the cask wall have maxima substantially less than the corresponding temperatures obtained when applying the steady-state regulatory boundary conditions. Therefore, it is certain that vital cask components and the spent fuel itself will not exceed the temperatures calculated by use of the steady-state interpretation of the 10CFR71 normal conditions

  3. A comparison of spent fuel shipping cask response to 10 CFR 71 normal conditions and realistic hot day extremes

    Energy Technology Data Exchange (ETDEWEB)

    Manson, S.J. [Texas Univ., Austin, TX (United States). Coll. of Engineering; Gianoulakis, S.E. [Sandia National Labs., Albuquerque, NM (United States)

    1994-02-01

    The structural properties of spent nuclear fuel shipping containers vary as a function of the cask wall temperature. An analysis is performed to determine the effect of a realistic, though bounding, hot day environment on the thermal behavior of spent fuel shipping casks. These results are compared to those which develop under a steady-state application of the prescribed normal thermal conditions of 10CFR71. The completed analysis revealed that the majority of wall temperatures, for a wide variety of spent fuel shipping cask configurations, fall well below those predicted by using the steady-state application of the regulatory boundary conditions. It was found that maximum temperatures at the cask surface occasionally lie above temperatures predicted under the regulatory condition. This is due to the conservative assumptions present in the ambient conditions used. The analysis demonstrates that diurnal temperature variations which penetrate the cask wall have maxima substantially less than the corresponding temperatures obtained when applying the steady-state regulatory boundary conditions. Therefore, it is certain that vital cask components and the spent fuel itself will not exceed the temperatures calculated by use of the steady-state interpretation of the 10CFR71 normal conditions.

  4. Testing and analyses of the TN-24P PWR spent-fuel dry storage cask loaded with consolidated fuel

    Energy Technology Data Exchange (ETDEWEB)

    McKinnon, M A; Michener, T E; Jensen, M F; Rodman, G R

    1989-02-01

    A performance test of a Transnuclear, Inc. TN-24P storage cask configured for pressurized water reactor (PWR) spent fuel was performed. The work was performed by the Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) for the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) and the Electric Power Research Institute. The performance test consisted of loading the TN-24P cask with 24 canisters of consolidated PWR spent fuel from Virginia Power's Surry and Florida Power and Light's Turkey Point reactors. Cask surface and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in both vertical and horizontal cask orientations. Transnuclear, Inc., arranged to have a partially insulated run added to the end of the test to simulate impact limiters. Limited spent fuel integrity data were also obtained. From both heat transfer and shielding perspectives, the TN-24P cask with minor refinements can be effectively implemented at reactor sites and central storage facilities for safe storage of unconsolidated and consolidated spent fuel. 35 refs., 93 figs., 17 tabs.

  5. Low-cost concepts for dry transfer of spent fuel and waste between storage and transportation casks

    International Nuclear Information System (INIS)

    The federal government may provide interim storage for spent fuel from commercial nuclear power reactors that have used up their available storage capacity. One of the leading candidate concepts for this interim storage is to place spent fuel in large metal shielding casks. The Federal Interim Storage (FIS) site may not have the capability to transfer spent fuel from transportation casks to storage casks and vice versa. Thus, there may be an incentive to construct a relatively inexpensive but reliable intercask transfer system for use at an FIS site. This report documents the results of a preliminary study of preconceptual design and analysis of four intercask transfer concepts. The four concepts are: a large shielded cylindrical turntable that contains an integral fuel handling machine (turntable concept); a shielded fuel handling machine under which shipping and storage casks are moved horizontally (shuttle concept); a small hot cell containing equipment for transferring fuel betwee shipping and storage casks (that enter and leave the cell on carts) in a bifurcated trench (trench concept) and a large hot cell, shielded by an earthen berm, that houses equipment for handling fuel between casks that enter and leave the cell on a single cart (igloo concept). Information derived for each of the concepts is operating, capital and relocation costs; implementation and relocation time requirements; and overall characteristics

  6. Low-cost concepts for dry transfer of spent fuel and waste between storage and transportation casks

    International Nuclear Information System (INIS)

    The federal government may provide interim storage for spent fuel from commercial nuclear power reactors that have used up their available storage capacity. One of the leading candidate concepts for this interim storage is to place spent fuel in large metal shielding casks. The Federal Interim Storage (FIS) site may not have the capability to transfer spent fuel from transportation casks to storage casks and vice versa. Thus, there may be an incentive to construct a relatively inexpensive but reliable intercask transfer system for use at an FIS site. This report documents the results of a preliminary study of preconceptual design and analysis of four intercask transfer concepts. The four concepts are: (1) a large shielded cylindrical turntable that contains an integral fuel handling machine (turntable concept); (2) a shielded fuel handling machine under which shipping and storage casks are moved horizontally (shuttle concept); (3) a small hot cell containing equipment for transferring fuel between shipping and storage casks (that enter and leave the cell on carts) in a bifurcated trench (trench concept); and (4) a large hot cell, shielded by an earthen berm, that houses equipment for handling fuel between casks that enter and leave the cell on a single cart (igloo concept). Information derived for each of the concepts is operating, capital and relocation costs; implementation and relocation time requirements; and overall characteristics. 1 reference, 4 figures, 2 tables

  7. Experimental investigation of heat removal performance of a concrete storage cask

    International Nuclear Information System (INIS)

    Highlights: • Thermal tests were performed to evaluate the heat removal performance of the concrete storage cask. • Passive heat removal system was well designed and worked adequately. • Half-blockage of the inlet has a relatively small effect. • Thermal integrity of the concrete is maintained under accident conditions. - Abstract: Spent nuclear fuel generated at nuclear power plants must be safely stored during interim storage periods. A concrete storage cask to safely store spent nuclear fuel should be able to adequately emit the decay heat from the spent nuclear fuel. Moreover, the concrete storage cask must ensure that the temperatures of the spent nuclear fuel assemblies are maintained within the allowable values for normal, off-normal, and accident conditions. Therefore, the concrete storage cask must be designed to have heat removal capabilities with appropriate reliability. However, the thermal conductivity of concrete is not good and the allowable temperature of concrete is lower than that of steel. In this study, a thermal test was performed to evaluate the heat removal performance of the concrete storage cask under development by KORAD (Korea Radioactive Waste Agency), under normal and off-normal conditions. In addition, a thermal test was performed to evaluate the thermal integrity of the concrete under accident conditions. The heat transfer rate to the ambient atmosphere by convective airflow through the passive heat removal system of the concrete storage cask was found to reach 93.5% under normal conditions. Thus, it was confirmed that the passive heat removal system was well designed and worked adequately. In addition, the heat transfer rate to the ambient atmosphere by convective airflow through the passive heat removal system under off-normal conditions was estimated to reach 87.4%. Therefore, it was deduced that the half-blockage of the inlet openings has a relatively small effect on the maximum temperatures and temperature distributions

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

    International Nuclear Information System (INIS)

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

  9. Regulators experiences in licensing and inspection of dry cask storage facilities

    International Nuclear Information System (INIS)

    Full text: All operating nuclear power reactors in the United States (US) are storing spent fuel in NRC licensed on-site spent fuel pools (SFPs). Most reactors were not designed to store, in these pools, the full amount of spent fuel generated during the life of plant operation. Utilities originally planned for spent fuel to remain in the SFPs for a few years after discharge from the reactor core and then to be sent to a reprocessing facility. However, the US Government declared a moratorium on reprocessing in 1977. Although the ban was later lifted, reprocessing has not been pursued as a feasible option. Consequently, utilities expanded the storage capacity of SFPs by the use of high-density storage racks. Eventually, utilities needed additional storage capacity. In response to these needs, NRC provided a regulatory alternative for interim spent fuel storage in dry cask storage systems. For spent fuel management, both pool storage and dry storage are safe methods, but there are significant differences. Pool storage requires a greater operational vigilance on the part of the nuclear power plant to maintain the performance of electrical and mechanical systems using pumps, piping and instrumentation. Dry storage technology uses passive cooling systems with robust cask designs requiring minimal operational vigilance. The United States Nuclear Regulatory Commission (NRC), through the combination of a rigorous licensing and inspection program, ensures the safety and security of dry cask storage. NRC authorizes the storage of spent fuel at an independent spent fuel storage installation (ISFSI) under two licensing options: site-specific licensing and general licensing. In July 1986, the NRC issued the first site-specific license to the Surry Nuclear Power Plant in Virginia, authorizing the interim storage of spent fuel in a dry storage cask configuration. Today, there are over 30 ISFSIs currently licensed by the NRC with over 700 loaded dry casks. Current projections

  10. ANALISIS CITRA PERMUKAAN THERMOCHROMIC LIQUID CRYSTAL BERDASARKAN NILAI STATISTIK HUE

    Directory of Open Access Journals (Sweden)

    Risti Suryantari

    2014-05-01

    Full Text Available Tujuan dari penelitian ini adalah untuk mengamati perbedaan citra pada permukaan Thermocromic Liquid Crystal (TLC R25C5W akibat perubahan temperatur. Metode yang digunakan untuk analisis adalah pengolahan citra berbasis morfologi matematika menggunakan perangkat lunak Matlab2013a. Citra asli yang diperoleh dalam bentuk RGB dikonversi menjadi HSV (hue, saturation, value, dengan mengambil komponen hue saja. Proses utama yang digunakan adalah opening dengan struktur elemen (SE line. Berdasarkan analisis visual pada citra akhir hasil pengolahan citra, tampak bahwa terdapat perbedaan setiap citra untuk berbagai temperatur berdasarkan tingkat kecerahannya. Secara kuantitatif perbedaan tersebut dapat dilihat dari nilai statistiknya. Nilai max dan mean citra hue semakin meningkat seiring dengan meningkatnya temperatur.

  11. ANALISIS KEBIJAKAN PENANAMAN MODAL ASING DI KABUPATEN BANTAENG

    OpenAIRE

    RACHMAT, ANUGRAH

    2015-01-01

    2015 Anugrah Rachmat, Nomor Induk E12110255,Program Studi Ilmu Pemerintahan, Fakultas Ilmu Sosial dan Ilmu Politik, Unversitas Hasanuddin, Menyusun Skripsi dengan judul ??? Analisis Kebijakan Penanaman Modal Asing di Kabupaten Bantaeng???. Di bawah bimbingan Dr. H. Andi Gau Kadir, M.A sebagai pemimbing I dan Dr. Hj.Nurlinah, M.Si sebagai pembimbing II. Penelitian ini bertujuan untuk mengambarkan model invetasi penanaman modal asing serta mengetahui prospeknya di Kabup...

  12. ANALISIS CITRA PERMUKAAN THERMOCHROMIC LIQUID CRYSTAL BERDASARKAN NILAI STATISTIK HUE

    OpenAIRE

    Risti Suryantari; - Flaviana

    2014-01-01

    Tujuan dari penelitian ini adalah untuk mengamati perbedaan citra pada permukaan Thermocromic Liquid Crystal (TLC) R25C5W akibat perubahan temperatur. Metode yang digunakan untuk analisis adalah pengolahan citra berbasis morfologi matematika menggunakan perangkat lunak Matlab2013a. Citra asli yang diperoleh dalam bentuk RGB dikonversi menjadi HSV (hue, saturation, value), dengan mengambil komponen hue saja. Proses utama yang digunakan adalah opening dengan struktur elemen (SE) line. Berdasark...

  13. Analisis Faktor – Faktor yang mempengaruhi Pertumbuhan Ekonomi Singapura

    OpenAIRE

    Sitepu, Wilsa Road Betterment

    2012-01-01

    The thesis investigate the factors predisposing economy growth in Singapore Country with using SEM (Structural Equation Model) consisting of factor analysis, path analysis, and regression. The research can be seen from its conceptual frame included in path analyze and data processing variable through AMOS (Analisis of Moment Structures) program as the multivariate analysis with some variables. The used data was secondary data taken from annual data and interpolated per t...

  14. MODEL TES DAN ANALISIS PRESTASI BELAJAR MATEMATIK SISWA SEKOLAH DASAR

    OpenAIRE

    Zamsir Zamsir

    2013-01-01

    Penelitian ini bertujuan untuk menemukan model tes dan analisis prestasi belajar siswa yang dapat dipakai untuk melakukan identifikasi level kemampuan dan menyusun profil pencapaian kompetensi siswa, khususnya di sekolah dasar. Penelitian ini merupakan penelitian pengembangan model yang dikembangkan menyangkut dua hal, yaitu: (1) prosedur dan langkah-langkah penyusunan tes serta teknik identifikasi level kemampuan siswa, dan (2) pelaporan hasil tes. Identifikasi dilakukan dengan cara menempat...

  15. ANALISIS PELAYANAN KB MANDIRI WANITA USIA SUBUR BERDASARKAN STATUS EKONOMI

    Directory of Open Access Journals (Sweden)

    Selma Siahaan

    2013-08-01

    Full Text Available Latar belakang: Penggunaan metode keluarga berencana (KB oleh Wanita Usia Subur (WUS antara lain dipengaruhi oleh faktor sosial ekonomi. Maka, masalah biaya yang harus dibayar oleh WUS untuk memperoleh pelayanan KB perludi pahami guna keberhasilan program KB Nasional. Metode: Studi menggunakan data Riset Kesehatan Dasar (Riskesdas 2010, bertujuan untuk mengetahui biaya pelayanan KB yang dibayar oleh perempuan usia 10–59 tahun (WUS yang pernah kawin menurut metode KB berdasarkan status ekonomi yang bersangkutan. Analisis data menggunakan analisis deskriptif, analisis bivariat dengan uji statistik chi-square. Hasil: Alat/obat KB disposible terbanyak yang digunakan WUS dari semua golongan status ekonomi adalah metode KB suntik, sedangkan kondom terbanyak digunakan oleh WUS dengan status ekonomi atas banyak. Sekitar 50% WUS dengan status ekonomi (SE rendah membayar biaya pelayanan KB berkisar Rp.10.000–Rp.15.000. Sementara itu WUS dengan SE atas terbanyak membayar sekitar > Rp. 20.000. Terdapat korelasi signifikan antara metode KB dengan biaya dan status ekonomi. Kesimpulan: Metode suntik masih merupakan metode pilihan bagi WUS di Indonesia dan pemilihan metode KB oleh WUS berhubungan dengan status ekonomi mereka. Oleh karena itu promosi kesehatan perlu terus digalakkan agar program KB pemerintah dengan IUD dapat lebih dikenal dan dapat diterima masyarakat. Di samping itu, diperlukan kebijakan khusus (bantuan kepada WUS dengan status ekonomi rendah sehingga IUD dapat diperoleh secara gratis atau setidaknya dengan biaya yang terjangkau.

  16. Analisis Strategi Peningkatan Layanan Sertifikasi Perangkat Telepon Seluler

    Directory of Open Access Journals (Sweden)

    Widya Budi Andhini

    2015-03-01

    Full Text Available Standardisasi sebagai unsur penunjang pembangunan mempunyai peran penting dalam usaha optimasi pendayagunaan sumber daya.Peningkatan standarisasi dan sertifikasi perangkat telekomunikasi dilaksanakan secara aktif dalam menjaga terlaksananya interkoneksi, seperti dalam Renstra KemKominfo Tahun 2010-2014. Tujuan kegiatan standardisasi telekomunikasi antara lain menjamin interoperabilitas dan interkonektivitas, mengendalikan mutu perangkat. Untuk meningkatkan penyelenggaraan sertifikasi perangkat, dilakukan kajian yang menggambarkan analisis strategi untuk meningkatkan penyelenggaraan sertifikasi telepon seluler. Penelitian ini menggunakan pendekatan kualitatif, dengan format desain deskriptif kualitatif, pendekatan kualitatif matriks SWOT untuk menghasilkan strategi. Pengumpulan data dilakukan melalui wawancara kepada narasumber terdiri dari Direktorat Standardisasi Ditjen SDPPI, Balai Besar Pengujian Perangkat Telekomunikasi, Telkom R&D Center, kuesioner kepada pabrikan dan toko handphone. Teknis analisis data menggunakan Model Miles and Huberman, dan analisis SWOT yaitu mengidentifikasi dan menganalisis faktor Strength (kekuatan, Weakness (Kelemahan, Opportunity (Peluang dan Threat (Tantangan. Hasil pembahasan memperlihatkan strategi masing-masing kolom di matriks SWOT adalah comparative advantage meliputi Roadmap pengembangan lembaga uji, perangkat, SDM; knowledge sharing; mobilization meliputi penyusunan aturan persyaratan teknis perangkat telekomunikasi, sosialisasi, survey; Divestment/ investment : facility sharing, kompetensi SDM, sosialisasi aplikasi online; dan Damage Control meliputi penambahan SDM, uji fungsi/pretest, pertemuan rutin antara Ditstand dan Lembaga Uji.

  17. Rilievo tridimensionale e analisi dei dissesti della Pieve di Romena

    Directory of Open Access Journals (Sweden)

    Stefano Bertocci

    2015-01-01

    Full Text Available La Pieve di San Pietro a Romena si trova nel comunedi Pratovecchio, nella vallata casentinese. La sua edificazionerisale, basandosi su quanto scolpito su di uncapitello, al 1152 d.C. per volere di Matilde di Canossa,si tratta quindi di un edificio romanico, anche se dell’edificiooriginario in realtà non resta molto, anche a causadei molti terremoti che hanno interessato la vallata.Sulla base di quanto richiesto dal DM 01/2008 sulle costruzionil’analisi del monumento è stata fatta seguendovari step conoscitivi: una corretta analisi delle fonti storiched’archivio per condurre le prime ipotesi di sviluppodell’edificio, un rilievo laser scanner in modo da avereun rilievo molto affidabile. Sulla base della nuvola dipunti sono state ottenute tutte le sezioni e i prospettinecessari a descrivere l’edificio e a completare l’analisicon gli elaborati materici, dei fotopiani, dei degradi superficiali.Parallelamente si è portata avanti l’analisi stratigraficasulla pieve, utile a capire come l’edificio è statoconcepito e quali trasformazioni ha subito nel tempo, inmodo poi da impostare il progetto di restauro.

  18. Design and fabrication of the retube transfer cask for Bruce N.G.S

    International Nuclear Information System (INIS)

    The retubing of CANDU reactors is a complex process which involves the removal and disposal of highly activated and contaminated calandria tubes and pressure tubes. For Bruce 'A' N.G.S., old pressure tubes will be removed from the reactor by cutting them into three segments; two end fitting assemblies which measure up to 3.2 m in length, and one pressure tube segment which measures up to 6.3 m. in length. Calandria tubes are 6.2 m in length. The function of the retube transfer cask is to provide for shielded transfer of these components between the reactor face and the in-ground disposal facility. This paper describes the design and fabrication of this cask. (author) 1 tab., 7 figs

  19. Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

    Energy Technology Data Exchange (ETDEWEB)

    Romano, T.

    1997-09-29

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  20. Direct disposal of transport an storage casks - status of the actual considerations

    International Nuclear Information System (INIS)

    For the final disposal of spent fuel elements and radioactive wastes from the spent fuel reprocessing two different concepts exist. The self-shielding POLLUX casks were developed for final disposal of spent fuels in underground repositories (gallery storage). For the high-level waste from reprocessing plants the concept of borehole storage of vitrified coquilles BSK3 was developed. for both concepts fuel elements and structural parts are supposed to be separated in conditioning facilities. An alternative concept (projects DIREGT) aimed to avoid conditioning is based on the direct final storage of transport and storage casks of the type CASTOR registered V in boreholes. The concepts have to consider the transport in the underground facility; the safety against criticality has to be demonstrated. An appropriate manipulation technique is to be developed.

  1. Ductile iron cask with encapsulated uranium, tungsten or other dense metal shielding

    International Nuclear Information System (INIS)

    In a cask for the transportation and storage of radioactive materials, an improvement in the shielding means which achieves significant savings in weight and increases in payload by the use of pipes of depleted uranium, tungsten or other dense metal, encapsulating polyethylene cores, dispersed in two to four rows of concentric boreholes around the periphery of the cask body which is preferably made of ductile iron. Alternatively, rods or small balls of these same shielding materials, alone or in combination, are placed in these bore holes. The thickness, number and arrangement of these shielding pipes or rods is varied to provide optimum protection against the neutrons and gamma radiation emitted by the particular radioactive material being transported or stored. (author) 4 figs

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

    International Nuclear Information System (INIS)

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

  3. Dual Purpose Cask for Dry Storage of Research Reactor Spent Fuel in Latin America

    International Nuclear Information System (INIS)

    Since 2001 Brazilian researchers have participated in a regional initiative, with researchers from other Latin American countries whom operate research reactors, to improve the regional capability in the management of spent fuel elements from these reactors. A dual purpose cask for transport and storage was selected as the best option for the long term dry storage of this material, and a half-scale model was designed, built and tested. Although the model failed the tests, its overall performance was considered very satisfactory and design and constructive features were changed as a result of the tests. A new test sequence with the modified cask model was scheduled for the first quarter of 2010. (author)

  4. Thermal hydraulic and neutronic analysis of dry cask storage systems for spent nuclear fuels

    International Nuclear Information System (INIS)

    Interim spent fuel storage systems must provide for the safe receipt, handling, retrieval and storage of spent nuclear fuel before reprocessing or disposal. In the context of achieving these objectives, the following features of the design were taken into consideration for metal shielded type storage systems; to maintain fuel subcritical, to remove spent fuel residual heat, to provide for radiation protection. These features in the design of a dry cask storage system were analyzed by employing COBRA-SFS and SCALE4.4 (ORIGEN, XSDOSE, CSAS6 ) codes for normal operation of the system under study. In accordance with safety assurance limits of International Atomic Energy Authority (IAEA), appropriate designs for Dry Cask Storage Systems (DCSS) were reached for 33000, 45000, and 55000 MWd/t burnup values and 5 and 10 years of cooling periods for spent fuel to be stored (Table 1)

  5. Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

    International Nuclear Information System (INIS)

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required

  6. Performance of CASTOR{sup R} HAW Cask Cold Trials for Loading, Transport and Storage of HAW canisters

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsmeier, Marco; Vossnacke, Andre [GNS Gesellschaft fuer Nuklear-Service mbH, Hollestrasse 7A, D-45127 Essen (Germany)

    2008-07-01

    On the basis of reprocessing contracts, concluded between the German Nuclear Utilities (GNUs) and the reprocessing companies in France (AREVA NC) and the UK (Nuclear Decommissioning Authority), GNS has the task to return the resulting residues to Germany. The high active waste (HAW) residuals from nuclear fuel reprocessing are vitrified and filled into steel cans, the HAW canisters. According to reprocessing contracts the equivalent number of HAW canisters to heavy metals delivered has to be returned to the country of origin and stored at an interim storage facility where applicable. The GNS' CASTOR{sup R} HAW casks are designed and licensed to fulfil the requirements for transport and long-term storage of HAW canisters. The new cask type CASTOR{sup R} HAW28M is capable of storing 28 HAW canisters with a maximum thermal power of 56 kW in total. Prior to the first active cask loading at a reprocessing facility it is required to demonstrate all important handling steps with the CASTOR{sup R} HAW28M cask according to a specific and approved sequence plan (MAP). These cold trials have to be carried out at the cask loading plant and at the reception area of an interim storage facility in Gorleben (TBL-G), witnessed by the licensing authorities and their independent experts. At transhipment stations GNS performs internal trials to demonstrate safe handling. A brand-new, empty CASTOR{sup R} HAW28M cask has been shipped from the GNS cask assembly facility in Muelheim to the TBL-G for cold trials. With this cask, GNS has to demonstrate the transhipment of casks at the Dannenberg transfer station from rail to road, transport to and reception at the TBL-G as well as incoming dose rate and contamination measurements and preparation for storage. After removal of all shock absorbers with a cask specific handling frame, tilting operation and assembly of the secondary lid with a pressure sensor, the helium leak tightness and 'Block-mass' tests have to be carried out

  7. Considerations for Disposition of Dry Cask Storage System Materials at End of Storage System Life

    International Nuclear Information System (INIS)

    Dry cask storage systems are deployed at nuclear power plants for used nuclear fuel (UNF) storage when spent fuel pools reach their storage capacity and/or the plants are decommissioned. An important waste and materials disposition consideration arising from the increasing use of these systems is the management of the dry cask storage systems' materials after the UNF proceeds to disposition. Thermal analyses of repository design concepts currently under consideration internationally indicate that waste package sizes for the geologic media under consideration may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded into the dry storage canisters currently in use. In the United States, there are already over 1650 of these dry storage canisters deployed and approximately 200 canisters per year are being loaded at the current fleet of commercial nuclear power plants. There is about 10 cubic meters of material from each dry storage canister system that will need to be dispositioned. The concrete horizontal storage modules or vertical storage overpacks will need to be reused, re-purposed, recycled, or disposed of in some manner. The empty metal storage canister/cask would also have to be cleaned, and decontaminated for possible reuse or recycling or disposed of, likely as low-level radioactive waste. These material disposition options can have impacts of the overall used fuel management system costs. This paper will identify and explore some of the technical and interface considerations associated with managing the dry cask storage system materials. (authors)

  8. Incentives for the use of depleted uranium alloys as transport cask containment structure

    International Nuclear Information System (INIS)

    Radioactive material transport casks use either lead or depleted uranium (DU) as gamma-ray shielding material. Stainless steel is conventionally used for structural containment. If a DU alloy had sufficient properties to guarantee resistance to failure during both nominal use and accident conditions to serve the dual-role of shielding and containment, the use of other structure materials (i.e., stainless steel) could be reduced. (It is recognized that lead can play no structural role.) Significant reductions in cask weight and dimensions could then be achieved perhaps allowing an increase in payload. The mechanical response of depleted uranium has previously not been included in calculations intended to show that DU-shielded transport casks will maintain their containment function during all conditions. This paper describesa two-part study of depleted uranium alloys: First, the mechanical behavior of DU alloys was determined in order to extend the limited set of mechanical properties reported in the literature. The mechanical properties measured include the tensile behavior the impact energy. Fracture toughness testing was also performed to determine the sensitivity of DU alloys to brittle fracture. Fracture toughness is the inherent material property which quantifies the fracmm resistance of a material. Tensile strength and ductility are significant in terms of other failure modes, however, as win be discussed. These mechanical properties were then input into finite element calculations of cask response to loading conditions to quantify the potential for claiming structural credit for DU. (The term ''structural credit'' describes whether a material has adequate properties to allow it to assume a positive role in withstanding structural loadings.)

  9. Concept for an all-purpose transport, storage, and disposal cask for spent nuclear fuel management

    International Nuclear Information System (INIS)

    The Tennessee Valley Authority believes that taking a systems approach to overall integration of spent fuel management with respect to onsite storage and disposal is essential. Their studies show that development of an integrated dry cask system suitable for onsite storage, transportation, monitored retrievable offsite storage, and perhaps use as a disposal container in a geologic repository offers the potential of the lowest overall economic, environmental, and social cost related to spent fuel management. 5 figures, 4 tables

  10. Nuclear Criticality Safety Evaluation of the 9965, 9968, 9972, 9973, 9974, and 9975 Shipping Casks

    International Nuclear Information System (INIS)

    A Nuclear Criticality Safety Evaluation (NCSE) has been performed for the 9965, 9968, 9972, 9973, 9974, and 9975 SRS-designed shipping casks. This was done in support of the recertification effort for the 9965 and 9968, and the certification of the newly designed 9972-9975 series. The analysis supports the use of these packages as Fissile Class I for shipment of fissionable material from the SRS FB-Line, HB-Line, and from Lawrence Livermore national Laboratory. six different types of material were analyzed with varying Isotopic composition, of both oxide and metallic form. The mass limits required to support the fissile Class I rating for each of the envelopes are given in the Table below. These mass limits apply if DOE approves an exception as described in 10 CFR 71.55(c), such that water leakage into the primary containment vessel does not need to be considered in the criticality analysis. If this exception is not granted, the mass limits are lower than those shown below. this issue is discussed in detail in sections 5 and 6 of the report.One finding from this work is important enough to highlight in the abstract. The fire tests performed for this family of shipping casks indicates only minimal charring of the Celotex thermal insulation. Analysis of the casks with no Celotex insulation (assuming it has all burned away), results in values of k-eff that exceed 1.0. Therefore, the Celotex insulation must remain intact in order to guarantee sub criticality of the 9972-9975 family of shipping casks

  11. Nonlinear Ultrasonic Diagnosis and Prognosis of ASR Damage in Dry Cask Storage

    Energy Technology Data Exchange (ETDEWEB)

    Qu, Jianmin [Northwestern Univ., Evanston, IL (United States); Bazant, Zdenek [Northwestern Univ., Evanston, IL (United States); Jacobs, Laurence [Georgia Inst. of Technology, Atlanta, GA (United States); Guimaraes, Maria [Electrical Power Research Institute, Palo Alto, CA (United States)

    2015-11-30

    Alkali-silica reaction (ASR) is a deleterious chemical process that may occur in cement-based materials such as mortars and concretes, where the hydroxyl ions in the highly alkaline pore solution attack the siloxane groups in the siliceous minerals in the aggregates. The reaction produces a cross-linked alkali-silica gel. The ASR gel swells in the presence of water. Expansion of the gel results in cracking when the swelling-induced stress exceeds the fracture toughness of the concrete. As the ASR continues, cracks may grow and eventually coalesce, which results in reduced service life and a decrease safety of concrete structures. Since concrete is widely used as a critical structural component in dry cask storage of used nuclear fuels, ASR damage poses a significant threat to the sustainability of long term dry cask storage systems. Therefore, techniques for effectively detecting, managing and mitigating ASR damage are needed. Currently, there are no nondestructive methods to accurately detect ASR damage in existing concrete structures. The only current way of accurately assessing ASR damage is to drill a core from an existing structure, and conduct microscopy on this drilled cylindrical core. Clearly, such a practice is not applicable to dry cask storage systems. To meet these needs, this research is aimed at developing (1) a suite of nonlinear ultrasonic quantitative nondestructive evaluation (QNDE) techniques to characterize ASR damage, and (2) a physics-based model for ASR damage evolution using the QNDE data. Outcomes of this research will provide a nondestructive diagnostic tool to evaluate the extent of the ASR damage, and a prognostic tool to estimate the future reliability and safety of the concrete structures in dry cask storage systems

  12. Development of scaling laws of heat removal and CFD assessment in concrete cask air path

    International Nuclear Information System (INIS)

    Highlights: • Vertical concrete cask was studied for PWR spent fuel dry storage. • Scaling laws were derived for facilities between prototype and half-scale model. • Computational Fluid Dynamics analysis was performed with 3D mesh generation. • Thermal radiation was considered with heat conduction and natural convection. - Abstract: This study investigates heat transfer in a concrete cask such as one used at intermediate storage facilities of PWR spent fuels. Sufficient removal of decay heat is necessary not to damage fuel cladding that functions as a radioactive materials barrier. The experimental design parameters were derived in the half-scale model for the assessment of the design analysis methodology including a CFD tool. The scaling methodology was developed to design the half-scale model of the concrete cask in the spent fuel dry storage through scaling analysis. As one of the most important scaling laws, the requirement of similarity was selected for the temperature rise between the inlet and the exit in the air path. Based on the natural circulation in the channel, the scaling law was derived for total canister power maintaining the similarity of the temperature rise. Then, the temperature calculation and the flow analysis were performed in concrete cask facilities for the prototype and the half scale model using Computational Fluid Dynamics code. Through the CFD simulations, the similarity of the temperature rise was demonstrated well between the inlet and the exit, and the exit temperature was well maintained between the prototype and the half scale model. Also the scaling ratios of air mass flow rate and exit velocity obtained by the scaling analysis were in good agreement with those predicted by CFD analysis

  13. Localization of cask and plug remote handling system in ITER using multiple video cameras

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, João, E-mail: jftferreira@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Vale, Alberto [Instituto de Plasmas e Fusão Nuclear - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ribeiro, Isabel [Laboratório de Robótica e Sistemas em Engenharia e Ciência - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

    2013-10-15

    Highlights: ► Localization of cask and plug remote handling system with video cameras and markers. ► Video cameras already installed on the building for remote operators. ► Fiducial markers glued or painted on cask and plug remote handling system. ► Augmented reality contents on the video streaming as an aid for remote operators. ► Integration with other localization systems for enhanced robustness and precision. -- Abstract: The cask and plug remote handling system (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell building and the Tokamak building in ITER. Different CPRHS typologies will be autonomously guided following predefined trajectories. Therefore, the localization of any CPRHS in operation must be continuously known in real time to provide the feedback for the control system and also for the human supervision. This paper proposes a localization system that uses the video streaming captured by the multiple cameras already installed in the ITER scenario to estimate with precision the position and the orientation of any CPRHS. In addition, an augmented reality system can be implemented using the same video streaming and the libraries for the localization system. The proposed localization system was tested in a mock-up scenario with a scale 1:25 of the divertor level of Tokamak building.

  14. Estimation of integrity of cast-iron cask against impact due to free drop test, (1)

    International Nuclear Information System (INIS)

    Ductile cast iron is examined to use for shipping and storage cask from a economic point of view. However, ductile cast iron is considered to be a brittle material in general. Therefore, it is very important to estimate the integrity of cast iron cask against brittle failure due to impact load at 9 m drop test and 1 m derop test on to pin. So, the F.E.M. analysis which takes nonlinearity of materials into account and the estimation against brittle failure by the method which is proposed in this report were carried out. From the analysis, it is made clear that critical flaw depth (the minimum depth to initiate the brittle failure) is 21.1 mm and 13.1 mm in the case of 9 m drop test and 1 m drop test on to pin respectively. These flaw depth can be detected by ultrasonic test. Then, the cask is assured against brittle failure due to impact load at 9 m drop test and 1 m drop test on to pin. (author)

  15. A conceptual redesign of an Inter-Building Fuel Transfer Cask

    International Nuclear Information System (INIS)

    The Inter-Building Fuel Transfer Cask, referred to as the IBC, is a lead shielded cask for transporting subassemblies between buildings on the Argonne National Laboratory-West site near Idaho Falls, Idaho. The cask transports both newly fabricated and spent reactor subassemblies between the Experimental Breeder Reactor-II (EBR-II), the Fuel Cycle Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The IBC will play a key role in the Integral Fast Reactor (IFR) fuel recycling demonstration project. This report discusses a conceptual redesign of the IBC which has been performed. The objective of the conceptual design was to increase the passive heat removal capabilities, reduce the personnel radiation exposure and incorporate enhanced safety features into the design. The heat transfer, radiation and thermal-hydraulic properties of the IBC were analytically modelled to determine the principal factors controlling the desip. The scoping studies that were performed determined the vital physical characteristics (i.e., size, shielding, pumps, etc.) of the MC conceptual design

  16. Thermal analysis of spent fuel shipping cask for application of metalized fuel

    International Nuclear Information System (INIS)

    Thermal analysis of spent fuel shipping cask loaded with 4 spent PWR fuel assemblies has been carried out using the fluent code. And the temperature distribution of cask for application of 4 metalized fuels equivalent to 16 PWR fuels has been also calculated. Total decay heat from 4 spent PWR fuels and 4 metalized spent fuels are 2.2 kW and 4.4 kW, respectively. The calculated temperatures for 4 spent PWR fuels were compared with the proven data presented from the safety analysis report of shipping cask. It has good agreement between two results. The maximum fuel rod temperatures inside the canisters of square and hexagonal types are estimated to be 269 .deg. C and 212 .deg. C, respectively. Therefore, it is found that the hexagonal canister loaded with metalized fuel rods is more advantageous in aspect of thermal characteristics and storage efficiency. Fuel temperature in the cavity of helium gas for hexagonal canister is lower than the temperature for spent PWR fuel

  17. Alternate approaches to verifying the structural adequacy of the Defense High Level Waste Shipping Cask

    Energy Technology Data Exchange (ETDEWEB)

    Zimmer, A.; Koploy, M.

    1991-12-01

    In the early 1980s, the US Department of Energy/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense high level waste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as one that fully complies with all applicable DOE, Nuclear Regulatory Commission (NRC), and Department of Transportation (DOT) regulations. General Atomics (GA) designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Laboratories (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This topical report presents the results of the two analytical approaches and the model testing results. The purpose of this work is to show that there are two viable analytical alternatives to verify the structural adequacy of a Type B package and to obtain an NRC license. It addition, this data will help to support the future acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing.

  18. A Criticality Evaluation of the GBC-32 Dry Storage Cask in PWR Burnup Credit

    International Nuclear Information System (INIS)

    The current criticality safety evaluation assumes the only unirradiated fresh fuels with the maximum enrichment in a dry storage cask (DSC) for conservatism without consideration of the depletion of fissile nuclides and the generation of neutron-absorbing fission products. However, the large conservatism leads to the significant increase of the storage casks required. Thus, the application of burnup credit which takes credit for the reduction of reactivity resulted from fuel depletion can increase the capacity in storage casks. On the other hand, the burnup credit application introduces lots of complexity into a criticality safety analysis such as the accurate estimation of the isotopic inventories and the burnup of UNFs and the validation of the criticality calculation. The criticality evaluation with an effect of burnup credit was performed for the DSC of GBC-32 by using SCALE 6.1/STARBUCS. keff values were calculated as a function of burnup and cooling time for four initial enrichments of 2, 3, 4, and 5 wt. % 235U. The values were calculated for the burnup range of 0 to 60,000 MWD/MTU, in increments of 10,000 MWD/MTU, and for five cooling times of 0, 5, 10, 20, and 40 years

  19. Alternative Splicing of a Novel Inducible Exon Diversifies the CASK Guanylate Kinase Domain

    Directory of Open Access Journals (Sweden)

    Jill A. Dembowski

    2012-01-01

    Full Text Available Alternative pre-mRNA splicing has a major impact on cellular functions and development with the potential to fine-tune cellular localization, posttranslational modification, interaction properties, and expression levels of cognate proteins. The plasticity of regulation sets the stage for cells to adjust the relative levels of spliced mRNA isoforms in response to stress or stimulation. As part of an exon profiling analysis of mouse cortical neurons stimulated with high KCl to induce membrane depolarization, we detected a previously unrecognized exon (E24a of the CASK gene, which encodes for a conserved peptide insertion in the guanylate kinase interaction domain. Comparative sequence analysis shows that E24a appeared selectively in mammalian CASK genes as part of a >3,000 base pair intron insertion. We demonstrate that a combination of a naturally defective 5 splice site and negative regulation by several splicing factors, including SC35 (SRSF2 and ASF/SF2 (SRSF1, drives E24a skipping in most cell types. However, this negative regulation is countered with an observed increase in E24a inclusion after neuronal stimulation and NMDA receptor signaling. Taken together, E24a is typically a skipped exon, which awakens during neuronal stimulation with the potential to diversify the protein interaction properties of the CASK polypeptide.

  20. Application of a chemical ion exchange model to transport cask surface decontamination

    International Nuclear Information System (INIS)

    Radionuclide contamination of stainless steel surfaces occurs during submersion in a spent fuel storage pool, Subsequent release or desorption of these contaminants from a nuclear fuel transportation cask surface under varying environmental conditions occasionally results in the phenomenon known as contamination 'weeping'. Experiments have been conducted to determine the applicability of a chemical ion exchange model to characterise the problem of cask contamination and release. Surface charge characteristics of Cr2O3 and stainless steel (304) powders have been measured to determine the potential for ion exchange at metal oxide-aqueous interfaces. The solubility of Co and Cs electrolytes at varying pH and the absorption characteristics of these ions on Cr2O3 and stainless steel powders in aqueous slurries have been studied. Experiments show that Co ions do reversibly absorb on these powder surfaces and, more specifically, that absorption occurs in the nominal pH range (pH = 4-6) of a boric acid moderated spent fuel pool. Desorption has been demonstrated to occur at pH≤3. Cs+ ions also have been shown to have an affinity for these surfaces although the reversibility of Cs+ bonding by H+ ion exchange has not been fully demonstrated. These results have significant implications for effective decontamination and coating processes used on nuclear fuel transportation casks. (author)

  1. Application of a chemical ion exchange model to transport cask surface decontamination

    International Nuclear Information System (INIS)

    Radionuclide contamination of stainless steel surfaces occur during submersion in a spent fuel storage pool. Subsequent release or desorption of these contaminants from a nuclear fuel transportation cask surface under varying environmental conditions occasionally results in the phenomenon known as contamination ''weeping.'' Experiments have been conducted to determine the applicability of a chemical ion-exchange model to characterize the problem of cask contamination and release. Surface charge characteristics of Cr2O3 and stainless steel (304) powders have been measured to determine the potential for ion exchange at metal oxide -- aqueous interfaces. The solubility of Co and Cs electrolytes at varying pH and the absorption characteristics of these ions on Cr2O3 and stainless steel powders in aqueous slurries have been studied. Experiments show that Co ions do reversibly adsorb on these powder surfaces and, more specifically, that adsorption occurs in the nominal pH range (pH = 4--6) of a boric acid-moderated spent fuel pool. Desorption has been demonstrated to occur at pH ≤ 3. Cs ions also have been shown to have an affinity for these surfaces although the reversibility of Cs+ bonding by H+ ion exchange has not been fully demonstrated. These results have significant implications for effective decontamination and coating processes used on nuclear fuel transportation casks. 8 refs., 5 figs

  2. Flexible path optimization for the Cask and Plug Remote Handling System in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Vale, Alberto, E-mail: avale@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Fonte, Daniel; Valente, Filipe; Ferreira, João [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ribeiro, Isabel [Laboratório de Robótica e Sistemas em Engenharia e Ciência, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Gonzalez, Carmen [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain)

    2013-10-15

    Highlights: ► Complementary approach for path optimization named free roaming that takes full advantage of the rhombic like kinematics of the Cask and Plug Remote Handling System (CPRHS). ► Possibility to find trajectories not possible in the past using the line guidance developed in a previous work, in particular when moving the Cask Transfer System (CTS) beneath the pallet or in rescue missions. ► Methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. -- Abstract: The Cask and Plug Remote Handling System (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell Building and the Tokamak Building in ITER along pre-defined optimized trajectories. A first approach for CPRHS path optimization was previously proposed using line guidance as the navigation methodology to be adopted. This approach might not lead to feasible paths in new situations not considered during the previous work, as rescue operations. This paper addresses this problem by presenting a complementary approach for path optimization inspired in rigid body dynamics that takes full advantage of the rhombic like kinematics of the CPRHS. It also presents a methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. The results gathered from 500 optimized trajectories are summarized. Conclusions and open issues are presented and discussed.

  3. Onsite storage of spent nuclear fuel in metalic spent fuel storage casks

    International Nuclear Information System (INIS)

    Virginia Electric and Power Company (Vepco) owns and operates two nuclear power stations within its system: the North Anna Power Station located in Louisa County, Virginia; and the Surry Power Station located in Surry County, Virginia. Each of these power stations has two pressurized water reactor operating units which share a common spent fuel pool. Under the Nuclear Waste Policy Act of 1982, Vepco is responsible for providing interim spent fuel storage until availability of the Federal Repository. Vepco has studied a number of options and has developed a program to provide the required onsite interim spent fuel storage. Options considered by Vepco included reracking, pin consolidation, dry storage and construction of a new spent fuel pool to provide the increased spent fuel storage capacity required. Vepco has selected reracking at North Anna combined with dry storage in metal spent fuel storage casks at Surrey to provide the required onsite spent fuel storage. A dry cask storage facility design and license application were developed and the license application was submitted to the NRC in October, 1982. The selection of the option to use dry cask storage of spent fuel at Surry represents the first attempt to license dry storage of spent nuclear fuel in the United States. This storage option is expected to provide an effective option for utilities without adequate storage space in their existing spent fuel pools

  4. Flexible path optimization for the Cask and Plug Remote Handling System in ITER

    International Nuclear Information System (INIS)

    Highlights: ► Complementary approach for path optimization named free roaming that takes full advantage of the rhombic like kinematics of the Cask and Plug Remote Handling System (CPRHS). ► Possibility to find trajectories not possible in the past using the line guidance developed in a previous work, in particular when moving the Cask Transfer System (CTS) beneath the pallet or in rescue missions. ► Methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. -- Abstract: The Cask and Plug Remote Handling System (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell Building and the Tokamak Building in ITER along pre-defined optimized trajectories. A first approach for CPRHS path optimization was previously proposed using line guidance as the navigation methodology to be adopted. This approach might not lead to feasible paths in new situations not considered during the previous work, as rescue operations. This paper addresses this problem by presenting a complementary approach for path optimization inspired in rigid body dynamics that takes full advantage of the rhombic like kinematics of the CPRHS. It also presents a methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. The results gathered from 500 optimized trajectories are summarized. Conclusions and open issues are presented and discussed

  5. Localization of cask and plug remote handling system in ITER using multiple video cameras

    International Nuclear Information System (INIS)

    Highlights: ► Localization of cask and plug remote handling system with video cameras and markers. ► Video cameras already installed on the building for remote operators. ► Fiducial markers glued or painted on cask and plug remote handling system. ► Augmented reality contents on the video streaming as an aid for remote operators. ► Integration with other localization systems for enhanced robustness and precision. -- Abstract: The cask and plug remote handling system (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell building and the Tokamak building in ITER. Different CPRHS typologies will be autonomously guided following predefined trajectories. Therefore, the localization of any CPRHS in operation must be continuously known in real time to provide the feedback for the control system and also for the human supervision. This paper proposes a localization system that uses the video streaming captured by the multiple cameras already installed in the ITER scenario to estimate with precision the position and the orientation of any CPRHS. In addition, an augmented reality system can be implemented using the same video streaming and the libraries for the localization system. The proposed localization system was tested in a mock-up scenario with a scale 1:25 of the divertor level of Tokamak building

  6. Standard review plan for reviewing safety analysis reports for dry metallic spent fuel storage casks

    International Nuclear Information System (INIS)

    The Cask Standard Review Plan (CSRP) has been prepared as guidance to be used in the review of Cask Safety Analysis Reports (CSARs) for storage packages. The principal purpose of the CSRP is to assure the quality and uniformity of storage cask reviews and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The CSRP also sets forth solutions and approaches determined to be acceptable in the past by the NRC staff in dealing with a specific safety issue or safety-related design area. These solutions and approaches are presented in this form so that reviewers can take consistent and well-understood positions as the same safety issues arise in future cases. An applicant submitting a CSAR does not have to follow the solutions or approaches presented in the CSRP. However, applicants should recognize that the NRC staff has spent substantial time and effort in reviewing and developing their positions for the issues. A corresponding amount of time and effort will probably be required to review and accept new or different solutions and approaches

  7. Gamma-ray control of metal and concrete cask radiation protection

    International Nuclear Information System (INIS)

    Metal and concrete cask for durable storage and transportation of the spent fuel is equipped by the remote control device for verification of radiation protection in particular concrete defects. Operation tenet is irradiation of the cask wall by gamma-rays with an exposure rate at the surface chart design. Introduction In compliance with the requirements of national standards and regulations which are valid in nuclear power engineering and also IAEA recommendations, transportation packaging modules (TPM) for long-term storage and shipment of the spent nuclear fuel (SNF) have to ensure rated protection against ionizing radiation and withstand emergency impacts while preserving integrity of tightness system and radiation protection. Special Mechanical Engineering Design Office (SMEDO) has developed to manufacture such a module on the basis of metal and concrete cask (TPM MCC) for spent nuclear fuel of RBMK- 1000 reactors, NPS (nuclear-powered submarines) etc. In general, the structure of MCC may be presented as three coaxial steel shells the space between them being filled with high-density (4 and 3.5 g/cm3) concrete of high ductility and reinforced with composite grid of bars, clamps and rings. We have developed a procedure to control radiation protection (RP) of this cask, RP integrity checks after dynamic testing which simulate emergency situation during transportation. Test bench of γ-control was designed and constructed. The task included assessment of the cask radiation protection parameters, correlation of estimated and pilot data, cask manufacturing quality control as well as assessment of its body concrete filling uniformity and finally, confirmation of RP integrity under applied dynamic loads. RP γ-control method has been selected for this task-solving. This method is based on radiometry of cask walls by irradiation from a radioactive source. So successive investigation and gamma-ray flow detection of cask wall are being performed in order to determine

  8. Alternate approaches to verifying the structural adequacy of the Defense High Level Waste Shipping Cask

    International Nuclear Information System (INIS)

    In the early 1980s, the US Department of Energy/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense high level waste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as one that fully complies with all applicable DOE, Nuclear Regulatory Commission (NRC), and Department of Transportation (DOT) regulations. General Atomics (GA) designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Laboratories (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This topical report presents the results of the two analytical approaches and the model testing results. The purpose of this work is to show that there are two viable analytical alternatives to verify the structural adequacy of a Type B package and to obtain an NRC license. It addition, this data will help to support the future acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing

  9. Dynamic Response Analysis of Storage Cask Lid Structure Subjected to Lateral Impact Load of Aircraft Engine Crash

    Energy Technology Data Exchange (ETDEWEB)

    Almomania, Belal; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Lee, Sanghoon [Keimyung Univ., Daegu (Korea, Republic of)

    2015-10-15

    Several numerical methods and tests have been carried out to measure the capability of storage cask to withstand extreme impact loads. Testing methods are often constrained by cost, and difficulty in preparation for several impact conditions with different applied loads, and areas of impact. Instead, analytic method is an acceptable process that can easily apply different impact conditions for the evaluation of cask integrity. The aircraft engine impact is considered as one of the most critical impact accidents on the storage cask that significantly affects onto the lid closure system and may cause a considerable release of radioactive materials. This paper presents a method for evaluating the dynamic responses of one upper metal cask lid closure without impact limiters subjected to lateral impact of an aircraft engine with respect to variation of the impact velocity. An assessment method to predict damage response due to the lateral engine impact onto metal storage cask has been studied by using computer code LS-DYNA. The dynamic behavior of the lid movements was successfully calculated by utilizing a simplified finite element cask model, which showed a good agreement with the previous research. The simulation analyses results showed that no significant plastic deformation for bolts, lid, and the cask body. In this study, the lid opening and sliding displacements are considered as the major factors in initiating the leakage path. This analysis may be useful for evaluating the instantaneous leakage rates in a connection with the sliding and opening displacements between the lid and the flange to ensure that the radiological consequences caused by an aircraft engine crash accident during the storage phase are within the permissible level.

  10. Discussion of Available Methods to Support Reviews of Spent Fuel Storage Installation Cask Drop Evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Witte, M.

    2000-03-28

    Applicants seeking a Certificate of Compliance for an Independent Spent Fuel Storage Installation (ISFSI) cask must evaluate the consequences of a handling accident resulting in a drop or tip-over of the cask onto a concrete storage pad. As a result, analytical modeling approaches that might be used to evaluate the impact of cylindrical containers onto concrete pads are needed. One such approach, described and benchmarked in NUREG/CR-6608,{sup 1} consists of a dynamic finite element analysis using a concrete material model available in DYNA3D{sup 2} and in LS-DYNA,{sup 3} together with a method for post-processing the analysis results to calculate the deceleration of a solid steel billet when subjected to a drop or tip-over onto a concrete storage pad. The analysis approach described in NUREG/CR-6608 gives a good correlation of analysis and test results. The material model used for the concrete in the analyses in NUREG/CR-6608 is, however, somewhat troublesome to use, requiring a number of material constants which are difficult to obtain. Because of this a simpler approach, which adequately evaluates the impact of cylindrical containers onto concrete pads, is sought. Since finite element modeling of metals, and in particular carbon and stainless steel, is routinely and accurately accomplished with a number of finite element codes, the current task involves a literature search for and a discussion of available concrete models used in finite element codes. The goal is to find a balance between a concrete material model with a limited number of required material parameters which are readily obtainable, and a more complex model which is capable of accurately representing the complex behavior of the concrete storage pad under impact conditions. The purpose of this effort is to find the simplest possible way to analytically represent the storage cask deceleration during a cask tip-over or a cask drop onto a concrete storage pad. This report is divided into three sections

  11. New generation of CASTOR registered casks for high enriched, high burn-up fuel from German NPP

    International Nuclear Information System (INIS)

    Requirements for new cask designs for transport and long-term dry storage of spent fuel assemblies (FA) from LWR-reactors are based on both increased source terms of the LWR FA including MOX FA, as well as the demand for economical optimisation of decommissioning costs by increased cask capacities. For this, cask development is the challenge to create and establish cask designs that can accommodate more FA with higher source terms, each under fixed boundary conditions (i.e. transport requirements and limitations of the power plants as crane loads and/or fixed maximum dimensions). This task has been elaborated by working simultaneously on different development actions each focussed to improve the cask performance. In the following a brief summary will be presented to give an overview which developments and investigations have been and are still will be performed for development and safety analyses of the new CASTOR registered -designs under the main subjects: material investigation and qualification, component tests and verifications, detailed design analysis and not at least design verification

  12. Thermodynamic analysis of thermosolar plants; Analisis termodinamico de plantas termosolares

    Energy Technology Data Exchange (ETDEWEB)

    Munoz, Felipe; Rojas, Armando [UNAM, Mexico, D.F. (Mexico)

    2000-07-01

    In this work we obtained the mathematical expressions to calculate the concentration area for parabolic trough and central tower solar systems in thermosolar plants. Thermodynamic analysis considering 1{sup s}t and 2{sup n}d laws were made for gas and steam energy conversion cycles joined to thermosolar plants. Economical analysis were made too. In this way, higher thermal efficiencies 57.4, 58, 57 y 58.1% and 2{sup n}d law efficiencies 76.6, 77.5, 76, y 77.6% were found with combined cycle and central tower solar system. However, the cheaper configuration (773.5 USD/MWh), found by the economic analysis, was the parabolic trough solar system with steam cycle. [Spanish] En este trabajo se determinan las expresiones matematicas para obtener el area de concentracion requerida en plantas termosolares de canal parabolico y de torre central combinadas con ciclos de generacion convencionales. Se realiza el analisis termodinamico con base en la primera y segunda ley a cada ciclo y asi mismo se efectua el analisis economico de dichos arreglos. Se encuentra que las mayores eficiencias termicas 57.4, 58, 57 y 58.1% y las eficiencias de 2 ley: 76.6, 77.5, 76, y 77.6% se tienen con sistema de torre central y ciclo combinado. El analisis economico expresa que el sistema solar con concentradores de canal parabolico con ciclos de vapor presenta el menor costo total de generacion (773.5 USD/MWh).

  13. Intorno all'analisi sintattica della frase semplice in italiano /

    Directory of Open Access Journals (Sweden)

    Josip Jernej

    2004-12-01

    Full Text Available Secondo una lunga tradizione confermata anche da opere illustri come la Sin­ tassi italiana di Raffaello Fornaciari del 1881, gli autori delle grammatiche italiane, nel trattare la struttura della frase semplice, adottano una soluzione fortemente influenzata dalla semantica, con i cosiddetti complementi indiretti. Trattasi di un modello che si differenzia completamente da quello adottato nelle grammatiche delle altre grandi lingue europee, come i1francese, il tedesco, il russo, in cui l'analisi della frase semplice è impostata  su criteri  essenzialmente sintattici.

  14. Evaluation of stress corrosion cracking in aqueous solution neutron shield of transport/storage cask for spent fuel

    International Nuclear Information System (INIS)

    Experimental evaluation proved that no chloride induced stress corrosion cracking will occur on the metal cask which utilizes propylene glycol aqueous solution as neutron shield. Crevice corrosion, precursor of cracking, occurs at about 0.4V vs. 0.1M-KCl silver silver-chloride reference electrode in aqueous solution with chloride concentration of more than 5 times higher than limit value. On the other hand, the electrochemical potential (ECP) of cask material was 0.08V in air saturated aqueous solution. Since ECP is much smaller than the crevice corrosion potential below which no crevice corrosion is expected, the possibility is very small for chloride induced stress corrosion cracking to occur on the cask. (author)

  15. Database of refractories for explosive and fire resistant steel cask for packaging and transportation of radioactive and hazardous materials

    International Nuclear Information System (INIS)

    This paper contains the results of mechanical and thermophysical properties investigations of the dense and porous refractory concretes (silicate (building), chamotte (metallurgical), alumina, zirconia (including ceramics)). Porosities of these materials were 20 - 50 %. Compression strength, thermal conductivity, thermal expansion, heat capacity and operation temperature for this refractories are discussed. The split-Hopkinson bar method was used for investigation of the strain rate about 1000 sec-1. For damage assessment of the severe events connected with overheating of the metal and oxides contents of cask and terrorist attack by means of the anti-tank weapons to cask we discussed resistance of a zirconia ceramics(concrete) to melted mixture Zr, UO2, Fe2O3 and Monroe jet. Our results testify that the porous zirconia ceramics can use in the impact limiter system of casks under mechanical, thermal and chemical attacks. (authors)

  16. Interaction of cosmic ray muons with spent nuclear fuel dry casks and determination of lower detection limit

    Science.gov (United States)

    Chatzidakis, S.; Choi, C. K.; Tsoukalas, L. H.

    2016-08-01

    The potential non-proliferation monitoring of spent nuclear fuel sealed in dry casks interacting continuously with the naturally generated cosmic ray muons is investigated. Treatments on the muon RMS scattering angle by Moliere, Rossi-Greisen, Highland and, Lynch-Dahl were analyzed and compared with simplified Monte Carlo simulations. The Lynch-Dahl expression has the lowest error and appears to be appropriate when performing conceptual calculations for high-Z, thick targets such as dry casks. The GEANT4 Monte Carlo code was used to simulate dry casks with various fuel loadings and scattering variance estimates for each case were obtained. The scattering variance estimation was shown to be unbiased and using Chebyshev's inequality, it was found that 106 muons will provide estimates of the scattering variances that are within 1% of the true value at a 99% confidence level. These estimates were used as reference values to calculate scattering distributions and evaluate the asymptotic behavior for small variations on fuel loading. It is shown that the scattering distributions between a fully loaded dry cask and one with a fuel assembly missing initially overlap significantly but their distance eventually increases with increasing number of muons. One missing fuel assembly can be distinguished from a fully loaded cask with a small overlapping between the distributions which is the case of 100,000 muons. This indicates that the removal of a standard fuel assembly can be identified using muons providing that enough muons are collected. A Bayesian algorithm was developed to classify dry casks and provide a decision rule that minimizes the risk of making an incorrect decision. The algorithm performance was evaluated and the lower detection limit was determined.

  17. Validation and benchmarking of calculation methods for photon and neutron transport at cask configurations

    International Nuclear Information System (INIS)

    The reliability of calculation tools to evaluate and calculate dose rates appearing behind multi-layered shields is important with regard to the certification of transport and storage casks. Actual benchmark databases like SINBAD do not offer such configurations because they were developed for reactor and accelerator purposes. Due to this, a bench-mark-suite based on own experiments that contain dose rates measured in different distances and levels from a transport and storage cask and on a public benchmark to validate Monte-Carlo-transport-codes has been developed. The analysed and summarised experiments include a 60Co point-source located in a cylindrical cask, a 252Cf line-source shielded by iron and polyethylene (PE) and a bare 252Cf source moderated by PE in a concrete-labyrinth with different inserted shielding materials to quantify neutron streaming effects on measured dose rates. In detail not only MCNPTM (version 5.1.6) but also MAVRIC, included in the SCALE 6.1 package, have been compared for photon and neutron transport. Aiming at low deviations between calculation and measurement requires precise source term specification and exact measurements of the dose rates which have been evaluated carefully including known uncertainties. In MAVRIC different source-descriptions with respect to the group-structure of the nuclear data library are analysed for the calculation of gamma dose rates because the energy lines of 60Co can only be modelled in groups. In total the comparison shows that MCNPTM fits very wall to the measurements within up to two standard deviations and that MAVRIC behaves similarly under the prerequisite that the source-model can be optimized. (author)

  18. A conceptual redesign of an inter-building fuel transfer cask

    International Nuclear Information System (INIS)

    The Inter-Building Fuel Transfer Cask, referred to as the IBC, is a lead shielded cask for transporting subassemblies between buildings on the Argonne National Laboratory-West site near Idaho Falls, Idaho. The cask transports both newly fabricated and spent reactor subassemblies between the Experimental Breeder Reactor-2 (EBR-2), the Fuel Cycle Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The IBC will play a key role in the Integral Fast Reactor (IFR) fuel recycling demonstration project. The existing IBC technology, designed and fabricated in the late fifties, is outdated and is a source of personnel exposure at ANL-W. The current IBC system requires forced argon cooling and has extremely limited passive cooling capabilities due to existing design features. A conceptual redesign of the IBC has been performed. The objective of the conceptual design was to increase the passive heat removal capabilities, reduce the personnel radiation exposure and incorporate enhanced safety features into the design. The heat transfer, radiation and thermal-hydraulic properties of the IBC were analytically modeled to determine the principal factors controlling the design. The scoping studies that were performed determined the vital physical characteristics (i.e., size, shielding, pumps, etc.) of the IBC conceptual design. The conceptual design for the IBC allows subassemblies with up to 800 Watts of decay heat to be passively cooled, a significant increase over the existing system. The new design which incorporates better passive cooling mechanisms will prevent inadvertent damage to the subassembly during postulated loss-of-power and loss-of-flow accident scenarios. The new design also decreases the radiation hazard to personnel by having fewer external systems, a better shield plug design, and surfaces that are easier to decontaminate. The control and monitoring system will also be state-of-the-art technology

  19. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J. C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

  20. FACSIM/MRS-1: Cask receiving and consolidation model documentation and user's guide

    International Nuclear Information System (INIS)

    The Pacific Northwest Laboratory (PNL) has developed a stochastic computer model, FACSIM/MRS, to assist in assessing the operational performance of the Monitored Retrievable Storage (MRS) waste-handling facility. This report provides the documentation and user's guide for the component FACSIM/MRS-1, which is also referred to as the front-end model. The FACSIM/MRS-1 model simulates the MRS cask-receiving and spent-fuel consolidation activities. The results of the assessment of the operational performance of these activities are contained in a second report, FACSIM/MRS-1: Cask Receiving and Consolidation Performance Assessment (Lotz and Shay 1987). The model of MRS canister storage and shipping operations is presented in FACSIM/MRS-2: Storage and Shipping Model Documentation and User's Guide (Huber et al. 1987). The FACSIM/MRS model uses the commercially available FORTRAN-based SIMAN (SIMulation ANalysis language) simulation package (Pegden 1982). SIMAN provides a set of FORTRAN-coded commands, called block operations, which are used to build detailed models of continuous or discrete events that make up the operations of any process, such as the operation of an MRS facility. The FACSIM models were designed to run on either an IBM-PC or a VAX minicomputer. The FACSIM/MRS-1 model is flexible enough to collect statistics concerning almost any aspect of the cask receiving and consolidation operations of an MRS facility. The MRS model presently collects statistics on 51 quantities of interest during the simulation. SIMAN reports the statistics with two forms of output: a SIMAN simulation summary and an optional set of SIMAN output files containing data for use by more detailed post processors and report generators

  1. Analisis SWOT Pada Warung Bang Man Di Kawasan Waruug Kopi (Warkop) Harapan Medan

    OpenAIRE

    Ricka Alamia Z

    2008-01-01

    Ricka Alamia Z, 2006. Analisis SWOT Pada Warnng Bang Man di Kawasan Warung Kopi (Warkop) Harapan Medan. Di bawah bimbingan Dra. Marhaini, MS, Ibu Prof. Dr. Ritha F Dalimunthe, SE, MSi (Ketua Departemen Manajemen), ibu Frida Ramadini, SE, MM (Penguji I), Ibu Dra, Lisa Marlina, MSi (Penguji II). Penelitian ini bertujuan untuk menganalisis SWOT yang ada pada Warung Bang Man. Analisis SWOT merupakan penilaian lingkungan internal berupa kekuatan (strength) dan kelemahan (weakness) serta lingkungan...

  2. ANALISIS KOMPARATIF STATIK TERHADAP PERMINTAAN KREDIT PADA BANK SYARIAH DAN BANK KONVENSIONAL DI KOTA MAKASSAR

    OpenAIRE

    Hamzah, Anggriawan Pradana

    2012-01-01

    Permasalahan dalam penelitian ini adalah apakah PDRB, bagi hasil dan laju inflasi berpengaruh terhadap permintaan kredit pada Bank Syariah serta PDRB, suku bunga dan laju inflasi berpengaruh terhadap permintaan kredit pada Bank Konvensional, serta apakah ada perbedaan antara permintaan kredit Bank Syariah dengan Bank Konvensional. Untuk menjawab permasalahan tersebut maka digunakan metode analisis regresi linear sederhana. Hasil analisis yang dilakukan menunjukkan bahwa pengaruh PDRB terha...

  3. Unsur-unsur Humanisme dalam Novel Tabula Rasa Karya Ratih Kumala : Analisis Psikologi Sastra

    OpenAIRE

    Tarigan, Itana

    2011-01-01

    Penelitian ini berjudul Unsur-unsur Humanisme dalam Novel Tabula Rasa Karya Ratih Kumala : Analisis Psikologi Sastra. Penelitian ini bertujuan untuk mengetahui unsur-unsur humanisme. Untuk mencapai tujuan tersebut maka data dikumpulkan dari novel yang berjudul yang berjudul Tabula rasa dengan menggunakan metode membaca heuristik dan hermeneutik. Data yang digunakan yaitu data primer dan data sekunder. Data primer diambil dari bahan analisis yaitu novel Tabula Rasa dan data sekunder diambil ...

  4. Interface Issues Arising Between Storage and Transport for Storage Facilities Using Storage/Transport Dual Purpose Dry Metal Casks

    International Nuclear Information System (INIS)

    The dual purpose dry metal casks were developed as a low cost and reliable design to handle spent fuel safely, not only in relation to storage, but also transportation. One of its main advantages is to enhance worker protection against radiation while reducing possible direct manipulation of the spent fuel. In order to define regulation and the use of this type of casks, a traditional approach can be used, based on the study of every individual aspect. However a new type of approach is possible, called the “holistic approach”, taking into account the different aspects as a whole. (author)

  5. Recommendations on Fuel Parameters for Standard Technical Specifications for Spent Fuel Storage Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M.

    2001-03-08

    The U.S. Nuclear Regulatory Commission (NRC) is currently reviewing the technical specifications for spent fuel storage casks in an effort to develop standard technical specifications (STS) that define the allowable spent nuclear fuel (SNF) contents. One of the objectives of the review is to minimize the level of detail in the STS that define the acceptable fuel types. To support this initiative, this study has been performed to identify potential fuel specification parameters needed for criticality safety and radiation shielding analysis and rank their importance relative to a potential compromise of the margin of safety.

  6. Monte Carlo simulation of radiation streaming from a radioactive material shipping cask

    International Nuclear Information System (INIS)

    Simulated detection of gamma radiation streaming from a radioactive material shipping cask have been performed with the Monte Carlo codes MCNP4A and MORSE-SGC/S. Despite inherent difficulties in simulating deep penetration of radiation and streaming, the simulations have yielded results that agree within one order of magnitude with the radiation survey data, with reasonable statistics. These simulations have also provided insight into modeling radiation detection, notably on location and orientation of the radiation detector with respect to photon streaming paths, and on techniques used to reduce variance in the Monte Carlo calculations. 13 refs., 4 figs., 2 tabs

  7. Automated-biasing approach to Monte Carlo shipping-cask calculations

    International Nuclear Information System (INIS)

    Computer Sciences at Oak Ridge National Laboratory, under a contract with the Nuclear Regulatory Commission, has developed the SCALE system for performing standardized criticality, shielding, and heat transfer analyses of nuclear systems. During the early phase of shielding development in SCALE, it was established that Monte Carlo calculations of radiation levels exterior to a spent fuel shipping cask would be extremely expensive. This cost can be substantially reduced by proper biasing of the Monte Carlo histories. The purpose of this study is to develop and test an automated biasing procedure for the MORSE-SGC/S module of the SCALE system

  8. Scoping design analyses for optimized shipping casks containing 1-, 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    1983-01-01

    This report details many of the interrelated considerations involved in optimizing large Pb, Fe, or U-metal spent fuel shipping casks containing 1, 2, 3, 5, 7, or 10-year-old PWR fuel assemblies. Scoping analyses based on criticality, shielding, and heat transfer considerations indicate that some casks may be able to hold as many as 18 to 21 ten-year-old PWR fuel assemblies. In the criticality section, a new type of inherently subcritical fuel assembly separator is described which uses hollow, borated stainless-steel tubes in the wall-forming structure between the assemblies. In another section, details of many n/..gamma.. shielding optimization studies are presented, including the optimal n/..gamma.. design points and the actual shielding requirements for each type of cask as a function of the age of the spent fuel and the number of assemblies in the cask. Multigroup source terms based on ORIGEN2 calculations at these and other decay times are also included. Lastly, the numerical methods and experimental correlations used in the steady-state and transient heat transfer analyses are fully documented, as are pertinent aspects of the SCOPE code for Shipping Cask Optimization and Parametric Evaluation. (While only casks for square, intact PWR fuel assemblies were considered in this study, the SCOPE code may also be used to design and analyze casks containing canistered spent fuel or other waste material. An abbreviated input data guide is included as an appendix).

  9. A Study on the Radiation Source Effect to the Radiation Shielding Analysis for a Spent-Fuel Cask Design with Burnup-Credit

    International Nuclear Information System (INIS)

    The radiation shielding analysis for a Burnup-credit (BUC) cask designed under the management of Korea Radioactive Waste Management Corporation (KRMC) was performed to examine the contribution of each radiation source affecting dose rate distribution around the cask. Various radiation sources, which contain neutron and gamma-ray sources placed in active fuel region and the activation source, and imaginary nuclear fuel were all considered in the MCNP calculation model to realistically simulate the actual situations. It was found that the maximum external and surface dose rates of the spent fuel cask were satisfied with the domestic standards both in normal and accident conditions. In normal condition, the radiation dose rate distribution around the cask was mainly influenced by activation source (60 Co radioisotope); in another case, the neutron emitted in active fuel region contributed about 90% to external dose rate at 1m distance from side surface of the cask. Besides, the contribution level of activation source was dramatically increased to the dose rates in top and bottom regions of the cask. From this study, it was recognized that the detailed investigation on the radiation sources should be performed conservatively and accurately in the process of radiation shielding analysis for a BUC cask.

  10. Development of a conditioning system for the dual-purpose transport and storage cask for spent nuclear fuel from decommissioned Russian submarines

    International Nuclear Information System (INIS)

    Russia, stores large quantities of spent nuclear fuel (SNF) from submarine and ice-breaker nuclear powered naval vessels. This high-level radioactive material presents a significant threat to the Arctic and marine environments. Much of the SNF from decommissioned Russian nuclear submarines is stored either onboard the submarines or in floating storage vessels in Northwest and Far East Russia. Some of the SNF is damaged, stored in an unstable condition, or of a type that cannot currently be reprocessed. In many cases, the existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing all of this fuel from remote locations. Additional transport and storage options are required. Some of the existing storage facilities being used in Russia do not meet health and safety and physical security requirements. The U.S. has assisted Russia in the development of a new dual-purpose metal-concrete transport and storage cask (TUK-108/1) for their military SNF and assisted them in building several new facilities for off-loading submarine SNF and storing these TUK-108/1 casks. These efforts have reduced the technical, ecological, and security challenges for removal, handling, interim storage, and shipment of this submarine fuel. Currently, Russian licensing limits the storage period of the TUK-108/1 casks to no more than two years before the fuel must be shipped for reprocessing. In order to extend this licensed storage period, a system is required to condition the casks by removing residual water and creating an inert storage environment by backfilling the internal canisters with a noble gas such as argon. The U.S. has assisted Russia in the development of a mobile cask conditioning system for the TUK-108/1 cask. This new conditioning system allows the TUK 108/1 casks to be stored for up to five years after which the license may be considered for renewal for an additional five years or the fuel will be shipped to

  11. IMPACLIB: a material property data library for impact analysis of radioactive material transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-12-01

    The paper describes the structural data library and graphical program for impact and stress analyses of radioactive material transport casks. Four kinds of material data, structure steels, stainless steels, leads and woods are compiled. These materials are main structural elements of casks. Structural data such as, coefficient of thermal expansion, modulus of longitudinal elasticity, modulus of transverse elasticity, Poisson`s ratio and stress-strain relationship have been tabulated. Main features of IMPACLIB are as follows: (1) data have been tabulated against temperature or strain rate, (2) thirteen kinds of polynominal fitting for stress-strain curve are available, (3) it is capable of graphical representations for structural data and (4) the IMPACLIB is able to be used on not only main frame computers but also work stations (OS UNIX) and personal computers (OS Windows 3.1). In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides a user`s guide for computer program and input data for the IMPACLIB. (author)

  12. Impact Analyses and Tests of Concrete Overpacks of Spent Nuclear Fuel Storage Casks

    International Nuclear Information System (INIS)

    A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches, those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters

  13. Improvement of Operational Safety of Dual-Purpose Cask for SNF in Storage

    International Nuclear Information System (INIS)

    By now more than 100 dual-purpose packaging sets TUK-108/1 are in operation in the mode of interim storage and transportation of SNF from decommissioned nuclear powered submarines (NPSs). In accordance with certificate, spent fuel is stored in TUK-108/1 on the premises of plants involved in NPS dismantlement for 2 years, whereupon it is transported for processing to PO Mayak. At one Far Eastern plant Zvezda involved in NPS dismantlement there arose a complicated situation due to necessity to extend period of storage of SNF in TUK-108/1. To ensure safety over a longer period of storage of SNF in TUK-108/1 it is essential to modify conditions of storage by removing of residual water and filling the inner cavity of the cask with an inert gas. This report describes key issues of cask drying technology, justification of terms of dry storage of naval SNF in TUK- 108/1 and RBMK-1000 SNF in UKhK-109, design features of the mobile drying facility, results of operation of the pilot facility at the Far Eastern plant Zvezda and cold testing on the shop factory for Leningrad NPP. (author)

  14. IMPACLIB: a material property data library for impact analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    The paper describes the structural data library and graphical program for impact and stress analyses of radioactive material transport casks. Four kinds of material data, structure steels, stainless steels, leads and woods are compiled. These materials are main structural elements of casks. Structural data such as, coefficient of thermal expansion, modulus of longitudinal elasticity, modulus of transverse elasticity, Poisson's ratio and stress-strain relationship have been tabulated. Main features of IMPACLIB are as follows: (1) data have been tabulated against temperature or strain rate, (2) thirteen kinds of polynominal fitting for stress-strain curve are available, (3) it is capable of graphical representations for structural data and (4) the IMPACLIB is able to be used on not only main frame computers but also work stations (OS UNIX) and personal computers (OS Windows 3.1). In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides a user's guide for computer program and input data for the IMPACLIB. (author)

  15. High capacity cask (TN28V) and International Transport System for the return shipment of vitrified high activity wastes

    International Nuclear Information System (INIS)

    The reprocessing of spent fuel generates different kinds of wastes. Among them fission products and non fissile actinides represent 98% of the radioactivity; these wastes are separated, concentrated, mixed with molten glass and poured into stainless steel containers. For political reasons, it is necessary to return these vitrified high activity wastes to the foreign countries which have decided to have their spent fuel reprocessed in France. So the transport of vitrified waste is vital for both the reprocessor and the utilities that have trusted the reprocessor and this operation has to be securely performed to give satisfaction to all concerned particles. For that reason Cogema will control the whole transport activity from La Hague plants to the receiving facilities of the customers. Therefore cogema will be responsible of the transport whatever the cask type (transport or storage) and will subcontract the transport operation to experienced companies such as Transnucleaire, PNTL or NTL, who will act on behalf of Cogema. Cogema will be the owner of the transport casks while the storage casks will normally be owned by the customers. Both cask types will of course have to comply with the requirements of La Hague, as published by Cogema

  16. 75 FR 57841 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6, Confirmation of Effective...

    Science.gov (United States)

    2010-09-23

    ... RIN 3150-AI88 List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6, Confirmation of... 42292). This direct final rule amended the NRC's spent fuel storage regulations at 10 CFR 72.214 to... spent fuel assemblies (36 undamaged Exxon fuel assemblies and up to 32 damaged fuel cans (in...

  17. Comparative changes in color features and pigment composition of red wines aged in oak and cherry wood casks.

    Science.gov (United States)

    Chinnici, Fabio; Natali, Nadia; Sonni, Francesca; Bellachioma, Attilio; Riponi, Claudio

    2011-06-22

    The color features and the evolution of both the monomeric and the derived pigments of red wines aged in oak and cherry 225 L barriques have been investigated during a four months period. For cherry wood, the utilization of 1000 L casks was tested as well. The use of cherry casks resulted in a faster evolution of pigments with a rapid decline of monomeric anthocyanins and a quick augmentation formation of derived and polymeric compounds. At the end of the aging, wines stored in oak and cherry barriques lost, respectively, about 20% and 80% of the initial pigment amount, while in the 1000 L cherry casks, the same compounds diminished by about 60%. Ethyl-bridged adducts and vitisins were the main class of derivatives formed, representing up to 25% of the total pigment amount in the cherry aged samples. Color density augmented in both the oak and cherry wood aged samples, but the latter had the highest values of this parameter. Because of the highly oxidative behavior of the cherry barriques, the use of larger casks (e.g., 1000 L) is proposed in the case of prolonged aging times. PMID:21548629

  18. Storing the Spent Nuclear Fuel in Dry Casks Licensed for a Century as an Alternative to Recycling Solution

    Science.gov (United States)

    Milincic, Radovan

    2010-02-01

    Management of spent nuclear power reactor fuels is one of the most urgent problems in nuclear technology. Yearly production of new spent fuel is in the range of thousands of tons, topping a couple of hundred thousand tons of spent fuel already. This material is extremely radioactive and currently there is no adequate international policy, control or management regarding it. I propose here an intermediate term solution to this problem, which will be technologically and economically sustainable: interim spent-fuel storage as an alternative to reprocessing. The reprocessing inherently increases the net amount of the plutonium, which can be used for production of nuclear arms. Moreover, it is an expensive process with the net effect of producing different type of radioactive waste. In particular, the development of a dry cask for nuclear waste storage on site and transport, licensed for a period of hundred years would provide a significantly less expensive solution in the recent future, giving a needed relief to crowded spent-fuel storage pools. Currently in the U.S, NRC licenses existing storage casks for 20 years; and licenses for some of the dry cask storage facilities in the U.S. are about to expire. The extended life dry casks will provide sufficient intermediate period toward a more efficient and/or technologically advanced solution for spent fuel. )

  19. Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    International Nuclear Information System (INIS)

    The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs

  20. Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    None

    1990-02-01

    The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs.

  1. Actinide partitioning-transmutation program. V. Preconceptual designs and costs of partitioning facilities and shipping casks, Appendix 4. Final report

    International Nuclear Information System (INIS)

    This Appendix contains cost estimate documents for the Fuels Fabrication Plant Waste Treatment Facility. Plant costs are summarized by Code of Accounts and by Process Function. Costs contributing to each account are detailed. Process equipment costs are detailed for each Waste Treatment Process. Service utility costs are also summarized and detailed. Shipping cask costs are provided

  2. Actinide partitioning-transmutation program. V. Preconceptual designs and costs of partitioning facilities and shipping casks, Appendix 4. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    This Appendix contains cost estimate documents for the Fuels Fabrication Plant Waste Treatment Facility. Plant costs are summarized by Code of Accounts and by Process Function. Costs contributing to each account are detailed. Process equipment costs are detailed for each Waste Treatment Process. Service utility costs are also summarized and detailed. Shipping cask costs are provided.

  3. How will regional brewers and their cask beer brands survive and prosper in the face of market changes resulting from Government intervention?

    OpenAIRE

    Jespersen, Paul

    2007-01-01

    This project is a study of the current state of the cask beer market in England, focussing on one particular regional cask beer producer, "A". The project has been undertaken with two main objectives in mind. Firstly, the project sets out to determine the nature and extent of the impact of recent Government intervention in the cask beer market; in particular, the introduction in 2002 of Progressive Beer Duty, or Small Relief, and, in the same year, the revocation of the Beer Orders legis...

  4. ANALISIS RELEVANSI LULUSAN PERGURUAN TINGGI DENGAN DUNIA KERJA

    Directory of Open Access Journals (Sweden)

    Ali Muhson

    2012-04-01

    Full Text Available Abstract: A Relevance Analysis of University Graduates with World of Work. Education should be oriented to the competencies required by the workforce as a percentage of unemployment among the educated increase continuously. This study aims to examine the relevance of YSU Economic Education graduates. The study only focuses on the type of work and subjects taught. The subject of this study is the alumni of Economic Education Study Program. Sampling technique used is snowball sampling. Data collection technique using questionnaires and documentation while the technique of data analysis using descriptive analysis. The result suggests that the majority of the graduates find their first job as private a teacher, a private employee and a tutor, while current job of the most graduates are private teacher, private employee, and civil servant (teacher. The data shows that more than 50 percent of the graduates work in the education area. This implies that the relevance level based on the type of work is categorized as sufficient. Majority of the graduates teaches social science, economic, and entrepreneurship, hence it can be concluded that the relevance level based on the subjects taught is highly relevant.   Keyword: relevance of graduates, type of work, unemployment, employment     Abstrak: Analisis Relevansi Lulusan Perguruan Tinggi dengan Dunia Kerja. Pendidikan harus berorientasi pada kompetensi yang dibutuhkan oleh dunia kerja karena persentase penganggur di kalangan terdidik terus meningkat. Penelitian ini bertujuan untuk mengkaji tingkat relevansi (kesesuaian lulusan Pendidikan Ekonomi UNY. Kajian hanya diarahkan pada jenis pekerjaan dan mata pelajaran yang diampu. Penelitian ini mengambil subjek alumni Prodi Pendidikan Ekonomi dari berbagai angkatan. Teknik sampling yang digunakan adalah snowball sampling. Teknik pengumpulan data menggunakan angket dan dokumentasi sedangkan teknik analisis data menggunakan analisis deskriptif. Penelitian ini

  5. ANALISIS SPASIAL FAKTOR LINGKUNGAN DAN KEJADIAN DBD DI KABUPATEN DEMAK

    Directory of Open Access Journals (Sweden)

    Musyarifatun Farahiyah

    2014-05-01

    Full Text Available AbstractDengue Haemoragic Fever (DHF was still a public health problem in Demak District, Central Java. In the year of 2012, there were 483 DHF cases and 6 of them were died. Based on those cases, there was no clear information how those cases spread related to environmental and demographic factor. This research aimed to do spatial analysis of DHF incidence then it was related to House Index (HI, Container Index (CI, and the density of houses and occupant. It was an observational research using survey method. This Research took 150 samples of DHF cases. The site of DHF incidence was identified using Geographic Positioning System (GPS device, to gained its coordinat. Environmental and demographic data was collected based on secondary information from District Health Office of Demak. The colected data would be analyzed spatially using ArcView GIS software. This research showed that there was no association between environmental factors (HI, CI with the Incidence Rate (IR of DHF (p-value < 0,05. However, the level of HI and CI was 13,17% and 7,08% respectively. It was a potential condition for DHF spreading in the community. Spatial analysis indicated that Mranggen Sub-district had the most number of DHF incidence that the pattern of spreading covered all area of villages. The higher of population and house density, the higher of Incidence Rate of DHFKeywords : spatial analysis of DHF, environmental factors, DemakAbstrakDemam berdarah dengue (DBD masih merupakan permasalahan serius di Kabupaten Demak Pripinsi Jawa Tengah. Pada tahun 2012, Di Kabupaten Demak terdapat 483 kasus DBD dengan 6 kematian, namun belum diketahui bagaimana sebaran kasus tersebut secara spasial dikaitkan dengan faktor lingkungan dan demografi. Penelitian ini bertujuan untuk melakukan analisis spasial kejadian DBD yang dihubungkan dengan House Index (HI, Container Index (CI, kepadatan penduduk dan kepadatan rumah. Penelitian ini merupakan penelitian observasional dengan metode

  6. Aplikasi Metode Lean Six Sigma Untuk Usulan Improvisasi Lini Produksi Dengan Mempertimbangkan Faktor Lingkungan. Studi Kasus: Departemen GLS (General Lighting Services) PT. Philips Lighting Surabaya

    OpenAIRE

    Miftachul Arifin; Hari Supriyanto

    2012-01-01

    Departemen GLS (General Lighting Services) PT. Philips Lighting Surabaya merupakan produsen lampu pijar. Pada pelaksanaan proses produksinya, perusahaan menemui beberapa kendala yang terkait dengan waste. Analisis lean six sigma dengan menggunakan value stream mapping menunjukkan terjadi defect di mesin finishing dan waiting di mesin mounting. EHS waste juga muncul yang mengindikasikan adanya dampak terhadap lingkungan dan kesehatan serta keselamatan pekerja. Pencarian akar permasalahan dilak...

  7. Analisis Komparatif Biaya Hutang Sukuk Dan Obligasi Perusahaan

    Directory of Open Access Journals (Sweden)

    Burhanuddin Jiwandaru

    2014-03-01

    Full Text Available Penelitian ini bertujuan untuk mengetahui perbandingan biaya hutang sukuk dibandingkan dengan obligasi sebagai sumber dana bagi perusahaan. Objek pada penelitian ini adalah seluruh sukuk dan obligasi korporasi yang diterbitkan perusahaan-perusahaan di Indonesia sejak Oktober 2006 sampai Juli 2010. Metode yang dilakukan dalam penelitian ini ialah dengan membandingkan rata-rata biaya hutang sukuk dengan rata-rata biaya hutang obligasi pada perusahaan. Hasil analisis data menunjukan bahwa rata-rata biaya hutang dari seluruh sukuk hampir sama daripada rata-rata biaya hutang dari seluruh obligasi. Bedasarkan data penelitian juga terdapat 21 obligasi yang memiliki biaya hutang yang sama dengan biaya hutang sukuknya, dan hanya terdapat dua obligasi memiliki biaya hutang yang sedikit lebih tinggi dari biaya hutang sukuknya.JEL Classification : G10, G14, G30Keywords : Biaya Hutang, Obligasi, Sukuk

  8. ANALISIS FAKTOR-FAKTOR YANG MEMPENGARUHI INDEPENDENSI PENAMPILAN AKUNTAN PUBLIK

    Directory of Open Access Journals (Sweden)

    Ardiani Ika S.

    2012-03-01

    Full Text Available Tujuan dari penelitian ini adalah untuk menguji secara empiris faktor-faktor yang mempengaruhi independensi penampilan akuntan publik baik secara parsial dan simultan. Faktor-faktor tersebut adalah financial interests, hubungan bisnis dengan klien, pelayanan asuransi dan audit, hubungan antara klien atau yg diaudit dengan auditor, kompetisi antara Kantor Akuntan Publik (KAP, ukuran KAP dan audit fee. Populasi dalam penelitian ini adalah para auditor di akuntan publik di Semarang tahun 2009. Metode penyeleksian sampel yang digunakan adalah convenience sampling dan jumlah respondennya 35. Alat analisis yang digunakan adalah multiple regression analysis. Dimulai dengan analisis kuantitatif untuk mengetes validitas dan reabilitas. Kemudian dilanjutkan uji asumsi klasik termasuk normality, multicollinearity, and heteroscedastisity. Hasil menunjukkan bahwa 6 variabel yang sudah dikaji secara simultan dan partial tersebut mempunyai efek yang signifikan terhadap independensi penampilan akuntan publik dengan R square 0.749. The objective of this study is to test empirically the factors that influence the independence of public accountant appearance both partially and simultaneously. The factors cover financial interests, business relationships with clients, assurance services and audit services, the length of relationship between client or auditee and auditor, the competition between Public Accountan Offices (KAPs, the size of KAP, and audit fee. The population in this study was the auditors who worked in the public accountant’s office in Semarang in 2009. The sample selection method is a convenience sampling and the  number of  respondents are 35. The Analysis tools used is multiple regression analysis. It starts by applying quantitative analysis to test the validity and reliability. Then, it is continued by testing the classical assumptions which include normality, multicollinearity, and heteroscedastisity. The result shows that the six

  9. ANALISIS KENDALA POTENSIAL PENERAPAN TOTAL QUALITY MANAGEMENT HOTEL PLAZA SEMARANG

    Directory of Open Access Journals (Sweden)

    Hendrajaya -

    2010-03-01

    Full Text Available Penelitian ini bertujuan untuk menemukan faktor-faktor yang terbentuk dari kendala penerapan Total Quality Management (TQM, sebagai upaya untuk menganalisis masalah-masalah potensial dari penerapan TQM di Hotel Plaza, Semarang. Sampel dari penelitian ini sejumlah 65 karyawan Hotel Plaza, Semarang yang terdiri dari manajer dan staf. Teknik analisis menggunakan Faktor Analisis. Hasil penelitian menunjukkan, bahwa kendala potensial dari implementasi TQM dapat disimpulkan menjadi lima faktor, yaitu faktor inkonsistensi, visi-misi, transformasi budaya, manajerial dan manajemen kualitas. Penelitian ini memberikan rekomendasi bahwa peningkatan fungsi-fungsi korporat melalui optimalisasi kualitas sebagai strategi bisnis dan orientasi pada kepuasan konsumen dengan melibatkan seluruh anggota organisasi, akan menjadi dasar untuk memaksimalkan daya saing perusahaan dalam hal produk, layanan, SDM, lingkungan dan seluruh proses yang dapat mendukung upaya perbaikan produktivitas. This study aims to determine the constraint factors and to analyze the potential problems of TQM implementation of Plaza Semarang Hotel. The sample of 65 employees of TQM implementation of Plaza Semarang Hotel, consist of managers and staffs. Factor analysis is used to achieve the results of the potential constraint implementation of Total Quality Management which can be summarized into five factors, they are: the factor of inconsistencies, the factor of vision and mission, cultural transformation factors, managerial factors and the factors of quality management. In order to increase the corporate functions through the optimization of quality as a business strategy and to achieve customer satisfaction, this study suggests that all members of the organization should be involved by the company as on-going basis to maximize the competitiveness of the organization on products, services, human resources, environment and all processes that support productivity improvement.

  10. ENCAPAI SUMBER DAYA MANUSIA UNGGUL (ANALISIS KINERJA DAN KUALITAS PELAYANAN

    Directory of Open Access Journals (Sweden)

    Ketut Sudarma

    2012-03-01

    Full Text Available Penelitian ini  bertujuan untuk menganalisis pengaruh kemampuan individu, kepuasan kerja, komitmen organisasi terhadap kinerja dan kinerja terhadap kualitas pelayanan. Sampel penelitian berjumlah 76 orang. Teknik pengambilan sampel menggunakan proportional random sampling. Variabel penelitian terdiri dari kemampuan individu, kepuasan kerja, komitmen organisasi, dan kinerja serta kualitas pelayanan  Analisis data menggunakan regresi dua tahap. Hasil penelitian menunjukkan bahwa kemampuan individu, kepuasan kerja dan komitmen organisasi mempunyai pengaruh positif signifikan terhadap kinerja dan kinerja berpengaruh positif signifikan terhadap kualitas pelayanan. Ini berarti semakin meningkat kinerja, maka semakin meningkat juga kualitas pelayanan. Dari hasil analisis deskriptif persentase tampak secara rata-rata  semua variabel penelitian  menunjukkan kategori baik, namun pada aspek-aspek tertentu pada masing-masing variabel masih terdapat kekurangan. Hasil penelitian ini diharapkan dapat menjadi masukan bagi pimpinan dalam meningkatkan kinerja dan kualitas pelayanan. Upaya perbaikan dimulai dari peningkatan kemampuan melalui pendidikan dan latihan, monitoring pelaksanaan kerja secara rutin, menumbuhkan komitmen kerja dan perbaikan standar operasi prosedur.The aim of the study is toanalyze the influence of individual abilities, job satisfaction, organizational commitment to performanceand performance againstservice quality. This samplingmethod is using 76 samples by the proportionalrandom sampling techniques. Variables consisted of individual ability, job satisfaction, organizational commitment, and performance andservice qualitydata analysisusingtwo-stageregression. The results showedthat theability of individuals, job satisfaction and organizational commitment had asignificant positive effecton performance and the performance ofa significant positive effecton quality of service. It means the higher performance, the higher service

  11. Meta Analisis sa Pagsusuri ng Maiikling Kwento sa mga Tesis at Disertasyon

    Directory of Open Access Journals (Sweden)

    Regina I. Cuizon

    2014-10-01

    Full Text Available Anumang mga pagbabago na makikita sa mundo ay bunga ng pananaliksik. Ang meta-analisis ay isa sa mga bunga ng pananaliksik na ginagamit sa kasalukuyan bilang teknik upang malaman iba pang mga nagsulputang informasyon. Ito’y pag-aaral sa mga pag-aaral. Isang kritikal at sistematikong pagsusuri sa istruktura ng mga pag-aaral. Maging gabay sa mga gradwadong paaralan sa pagpili ng paksang pagaaralan. Pangunahing layunin na matiyak ang mga pamamaraan sa pagsusuri ng maiikling kwento ng mga tesis at disertasyong nagawa mula sa mga piling SUCs. Desinyong kwalitatibo - kontent analisis sa pagsusuri ng: kaanyuan ayon sa suliranin, metodolohiya, paglalahad at interpretasyon ng mga datos, natuklasan, konklusyon, at rekomendasyon; kahinatnan ayon sa pagkatulad, pagkakaiba at kabuluhan; Emerging tema. Napag-alaman na ang karaniwang pinag-aralan ay 30% kahalagahang pangkatauhan, 20% larawangdiwa ng mga kababaihan at 10% gramatikang aspeto. Sa metodolohiyang ginamit, 80% desinyong kwalitatibo at 20% kwantitatibo-kwalitatibo. Sa paglikom ng datos 60% diretsahang pagsusuri, 30% talatanungan at 10% tseklis. Sa pag-analisa 90% kontent analisis at 10% gramatikal analisis. Batay sa natuklasan, ang pagsusuri ng maiikling kwento sa mga tesis at disertasyon mula sa iba’t ibang paaralan gamit ang meta analisis ay isang epektibo, mabisa, objektibong paraan at kagamitan na magagamit sa makatarungang paghatol; pamumuna sa kabuluhan at kagandahan; paghaham- bing sa mga kritikal na isyu; at pormulasyon ng panibagong pamantayan at batas.

  12. Spent Nuclear Fuel (SNF) Project Cask and MCO Helium Purge System Design Review Completion Report Project A.5 and A.6

    Energy Technology Data Exchange (ETDEWEB)

    ARD, K.E.

    2000-04-19

    This report documents the results of the design verification performed on the Cask and Multiple Canister Over-pack (MCO) Helium Purge System. The helium purge system is part of the Spent Nuclear Fuel (SNF) Project Cask Loadout System (CLS) at 100K area. The design verification employed the ''Independent Review Method'' in accordance with Administrative Procedure (AP) EN-6-027-01.

  13. Experimental Study on Influence of Mechanical Vibration during Transport of Transport/Storage Cask for Spent Nuclear Fuel on Containment Performance of Metal Gasket during Storage in Japan

    International Nuclear Information System (INIS)

    Transport casks of spent nuclear fuel will receive mechanical vibration during transport. It is known that the containment performance of metal gaskets is influenced by large external load or displacement. Quantitative influence of such vibration during transport on the containment performance of the metal gasket has not been known, but is crucial information particularly if the cask is stored as it is after the transport

  14. Interim storage of spent fuel assemblies from VVER-reactors, taking as an example the cask dry storage for the Czech Dukovany Nuclear Power Plant

    International Nuclear Information System (INIS)

    The nuclear fuel cycle services from the former Soviet Union were fundamentally changed in 1989. The necessity of intermediate spent fuel storage increased in Czechoslovakia in short term. After performing an international comparison and request for proposals, Czechoslovakia represented by the electrical utility CEZ in Prague, decided in favor of a dry cask storage concept for the nuclear power plant Dukovany. The selection process among the offered solutions and the dry cask storage concept is discussed

  15. Direct final disposal of transport and storage casks. A realizable technical concept

    International Nuclear Information System (INIS)

    GNS and DBE TEC developed possible alternatives and supplementary concepts to the existing German reference concept POLLUX and the concept of direct final disposal in boreholes (BSK3) the concept of direct final disposal of transport and storage casks (DIREGT). Advantages of this include the avoidance of necessary elaborate segmentation of fuel elements and core structures, the reduction of waste package transfers and standardized technical equipment for the final disposal engineering. The tasks to be studied include the adaptation of the shaft lifting to the high workload, the adaptation of the underground hauling to the high loads and the development of an appropriate storage technology, considerations concerning the safety with respect to criticality for the demonstration of long-term safety. The basic feasibility of the concept has been demonstrated, the work to be done concerns the demonstration of approvability of the concept for licensing purposes.

  16. Shielding and Containment Evaluations of the NAC-LWT Cask with Tritium Burnable Poison Rods

    International Nuclear Information System (INIS)

    In 1989, the NAC legal weight truck cask (NAC-LWT) was approved by the U.S. Nuclear Regulatory Commission to transport either one pressurized water reactor (PWR) fuel assembly or two boiling water reactor (BWR) fuel assemblies. Since that time, license amendments have allowed the shipment of high-burnup PWR and BWR fuel rods, MTR-type research reactor fuel elements, and TRIGA-type fuel elements. In 1999, DOE approved an NAC-LWT submittal for a shipment of lead test assemblies (LTAs) containing tritium-producing burnable poison rods (TPBARs). This paper presents the 10 CFR Part 71 shielding and containment evaluations of the NAC-LWT with the LTA payload

  17. Preliminary investigation of aluminium foam as an energy absorber for nuclear transportation cask

    Energy Technology Data Exchange (ETDEWEB)

    Rajendran, R. [BARC Facilities, Bhabha Atomic Research Centre, Kalpakkam, Tamilnadu 603 102 (India)], E-mail: rajurajendr@yahoo.co.in; Prem Sai, K.; Chandrasekar, B. [BARC Facilities, Bhabha Atomic Research Centre, Kalpakkam, Tamilnadu 603 102 (India); Gokhale, A. [Defence Metallurgical Research Laboratory, Kanchanbagh, Hyderabad 500 058 (India); Basu, S. [BARC Facilities, Bhabha Atomic Research Centre, Kalpakkam, Tamilnadu 603 102 (India)

    2008-10-15

    Closed cell aluminum foam is investigated for its impact energy absorption characteristics. For this purpose, a drop hammer of 106 kg was fabricated. A free-fall drop tower was used for the experiments. The hammer was impacted on the rigid foundation with and without aluminium foam at its bottom. Acceleration-time history was recorded for each drop. Deflection of the foam undergoing impact was measured. Compression test was carried out on a foam cylinder to obtain the representative stress-strain diagram from which energy-deflection diagram was derived. Gibson-Ashby's plateau stress-density relation was applied to evaluate the energy-deflection characteristics of foams of different densities, which were eventually applied to the theoretical predictions. Force reduction factor offered by foam is attractive enough to considering, it as the candidate for sacrificial member of the transportation cask.

  18. Criticality safety evaluation for long term storage of FFTF fuel in interim storage casks

    International Nuclear Information System (INIS)

    It has been postulated that a degradation phenomenon, referred to as ''hot cell rot'', may affect irradiated FFTF mixed plutonium-uranium oxide (MOX) fuel during dry interim storage. ''Hot cell rot'' refers to a variety of phenomena that degrade fuel pin cladding during exposure to air and inert gas environments. It is thought to be a form of caustic stress corrosion cracking or environmentally assisted cracking. Here, a criticality safety analysis was performed to address the effect of the ''hot cell rot'' phenomenon on the long term storage of irradiated FFTF fuel in core component containers. The results show that seven FFTF fuel assemblies or six Ident-69 pin containers stored in core component containers within interim storage casks will remain safely subcritical

  19. U.S. Regulatory Research Program for Implementation of Burnup Credit in Transport Casks

    International Nuclear Information System (INIS)

    In 1999 the U.S. Nuclear Regulatory Commission (U.S. NRC) initiated a research program to support the development of technical bases and guidance that would facilitate the implementation of burnup credit into licensing activities for transport and dry cask storage. This paper reviews the following major areas of investigation: (1) specification of axial burnup profiles, (2) assumption on cooling time, (3) allowance for assemblies with fixed and removable neutron absorbers, (4) the need for a burnup margin for fuel with initial enrichments over 4 wt %, and (5) evaluation of assay data and critical experiments. The capabilities of a new computational tool that facilitates the performance and coupling of the depletion and criticality analyses needed for burnup credit are also discussed

  20. Development of dual purpose (storage and transport) metal casks in Spain

    International Nuclear Information System (INIS)

    The Spanish Nuclear Program consists of nine nuclear power plants with an overall capacity of 7.4 GWe, representing around 38% of the electricity share. In 1982 the National Energy Plan established the open cycle as the strategy to be followed thereby halting the reprocessing option. As a consequence of this Plan, a state-owned company, ENRESA, responsible to the ministry of Industry and energy, was created in 1984 with the responsibility to manage all kinds of radioactive wastes. To comply with its responsibilities for the interim storage of spent fuel, ENRESA has designed a strategy based on the use of dual purpose metal casks for the initial demand of additional storage capacity

  1. Quality assurance, fabrication, and accompanying quality control of CASTOR {sup registered} transport and storage casks

    Energy Technology Data Exchange (ETDEWEB)

    Beverungen, M.; Laug, R. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2004-07-01

    In Germany, the Federal Institute for Materials Research and Testing (BAM, Berlin) acts as the competent authority for the approval of quality assurance measures of packages used in the transport of radioactive material. For this purpose, the German Federal Ministry of Transport issued the ''Technical Guideline on Measures for Quality Assurance (QM) and Quality Surveillance (QUe) for packages for transport of radioactive materials (TRV 006)''. Due to this guideline, every applicant for a cask license is requested to issue a written programme for design, fabrication, testing, documentation, operation, maintenance, and recurrent testing of packages. This programme has to be approved by BAM. Therefore, the company implemented and maintains a programme on basis of the TRV 006 and a Quality Management System on basis of the ISO 9001:2000 which cover all requirements of the guideline to ensure a controlled processes. The system is regularly assessed by BAM.

  2. Application of surface complexation modeling to the understanding of transportation cask weeping

    Energy Technology Data Exchange (ETDEWEB)

    Granstaff, V.E.; Chambers, W.B.

    1993-11-01

    A new application for surface complexation modeling is described. These models, which describe chemical equilibria among aqueous and adsorbed species, have typically been used for predicting groundwater transport of contaminants by modeling the natural adsorbents as various metal oxides. We have shown that this type of modeling can also be used to explain stainless steel surface contamination and decontamination mechanisms. Stainless steel transportation casks that are submerged in a spent fuel storage pool at nuclear power stations, can become contaminated with radionuclides such as {sup 137}CS, {sup 134}Cs, and {sup 60}Co. Subsequent release or desorption of these contaminants under varying environmental conditions occasionally results in the phenomenon known as ``cask weeping.`` We have postulated that contaminants in the storage pool adsorb onto the hydrous metal oxide surface of the passivated stainless steel and are subsequently released during transportation, due to varying environmental factors, such as humidity, road salt, dirt, and acid rain. It is well known that 304 stainless steel has a chromium enriched passive surface layer; thus its adsorption behavior should be similar to that of chromium oxide. Presented here are adsorption data for Co{sup +2} on Cr{sup 2}O{sup 3} which simulate the stainless steel surface contamination. These data are interpreted using electrostatic surface complexation models. The FITEQL computer program was used to obtain the electrostatic model constants from the experimental data. Because the concentrations of contaminants in the storage pool are too low to be measured accurately by conventional chemical analysis techniques, MINTEQA2 can be used, with the fitted constants, to extrapolate the equilibria to the low concentrations representative of storage pool water.

  3. Study of a brazilian cask and its installation for PWR spent nuclear fuel dry storage

    International Nuclear Information System (INIS)

    Spent nuclear fuel (SNF) is removed from the nuclear reactor after the depletion on efficiency in generating energy. After the withdrawal from the reactor core, the SNF is temporarily stored in pools at the same site of the reactor. At this time, the generated heat and the short and medium lived radioactive elements decay to levels that allow removing SNF from the pool and sending it to temporary dry storage. In that phase, the fuel needs to be safely and efficiently stored, and then, it can be retrieved in a future, or can be disposed as radioactive waste. The amount of spent fuel increases annually and, in the next years, will still increase more, because of the construction of new nuclear plants. Today, the number of new facilities back up to levels of the 1970's, since it is greater than the amount of decommissioning in old installations. As no final decision on the back-end of the nuclear fuel cycle is foreseen in the near future in Brazil, either to recover the SNF or to consider it as radioactive waste, this material has to be isolated in some type of storage model existing around the world. In the present study it is shown that dry SNF storage is the best option. A national cask model for SNF as well these casks storage installation are proposed. It is a multidisciplinary study in which the engineering conceptual task was developed and may be applied to national SNF removed from the Brazilian power reactors, to be safely stored for a long time until the Brazilian authorities will decide about the site for final disposal. (author)

  4. Summary report on optimized designs for shipping casks containing 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel

    International Nuclear Information System (INIS)

    The purpose of this study was to develop new conceptual designs for large Pb, Fe, and U-shielded spent fuel casks which have been optimized for the shipment of 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel assemblies. Design specifications for about 100 cases of potential interest are presented along with a brief 20-page synopsis of the associated analyses. Optimized shielding requirements are presented for each type of cask as a function of the age of the spent fuel and the number of assemblies in the cask. With respect to criticality, a new type of inherently subcritical fuel assembly separator is described which uses hollow, borated stainless-steel tubes in the wall-forming structure between the assemblies. Steady-state and transient heat transfer analyses for casks under nominal and accident conditions were performed using the SCOPE code for Shipping Cask Optimization and Parametric Evaluation. Based on criticality, shielding, and heat transfer considerations, it appears that optimized cask designs could be developed to carry 15 to 18 five-year-old PWR fuel assemblies or as many as 18 to 21 ten-year-old PWR fuel assemblies. 4 figures, 4 tables

  5. CASTOR registered 1000/19. Shipping and storage cask for dry intermediate storage of spent fuel elements from the Temelin nuclear power station

    International Nuclear Information System (INIS)

    The CASTOR registered 1000/19-type cask newly developed by GNS Gesellschaft fuer Nuklear-Service mbH has been designed for shipment and dry storage of 19 spent fuel elements from the two VVER-1000 pressurized water reactors of the Temelin nuclear power station. The design and expert approval followed the internationally recognized shipping regulations of IAEA and the Czech rules and regulations for dry intermediate storage. One June 21, 2010, the Czech regulatory authority issued a type approval (integral transport and storage permit) valid for a period of 5 years for the casks of the new CASTOR registered 1000/19 line. As early as on September 9, 2010, which is less than 4 years after the delivery contract had been signed by CEZ and GNS in November 2006, the first cask of the CASTOR registered 1000/19 line was loaded with 19 spent fuel elements in only 5 days and emplaced in the newly built intermediate cask store. Loading, handling, in-plant transport, and emplacement of the cask made successful use of the extensive handling and management equipment furnished by GNS. After successful commissioning of the equipment and cold handling in both power plant units and in the intermediate cask store, successful loading and emplacement represents one of the most important milestones in the project. (orig.)

  6. Transportation of 33 irradiated MTR fuel assemblies from FRM/Garching to Savannah River Site, USA, using a GNS transport cask and using a new loading device

    Energy Technology Data Exchange (ETDEWEB)

    Dreesen, K.; Goetze, H.G.; Holst, L. [GNS, D-45127 Essen (Germany); Gerstenberg, H.; Schreckenbach, K. [Technical University of Munich, D-85748 Garching (Germany)

    2000-07-01

    According to the Department of Energy program of the return spent fuel from the foreign research reactors operators, 33 irradiated MTR box shaped fuel assemblies from the Technical University Munich were shipped to SRS/USA. The fuel assemblies were irradiated for typically 800 full days and, after a sufficient cooling time, loaded into a GNS 16 cask. The GNS 16 cask is a new transport cask for box shaped MTR fuel assemblies and TRIGA fuel assemblies and was used for the first time at the FRM Garching. The capacity of the cask is 33 box shaped MTR fuel assemblies. During the loading of the fuel assemblies, a newly developed loading device was used. The main components of the loading device are the transfer flask, the shielded loading lock, adapter plate and a mobile water tank. The loading device works mechanically with manpower. For the handling of the transfer flask, a crane with a capacity of 5 metric tons is necessary. During installation of the lid the mobile water pool is filled with demineralized water and the shielded loading passage is taken away. After that the lid is put on the cask. After drainage, the mobile water pool is disassembled, and the cask is dewatered. Finally leak tests of all seals are made. The achieved leakage rate was <1.0 x 10{sup -5} Pa x I/s. The work in FRM was done between 03.02.99 and 12.02.99 including a dry run and leak test. (author)

  7. Full-scale prototyping of the Hitachi dual-purpose metal cask and verification of its heat transfer characteristics

    International Nuclear Information System (INIS)

    Hitachi has been developing dual-purpose metal casks for transport and storage of spent nuclear fuels. The Hitachi cask, HDP69B can store 69 BWR fuel assemblies. The cask features are as follows. 1) The fuel basket is assembled mainly with plates of borated stainless steel. The plates are not welded, but cross-inserted into each other like the dividers in an egg carton. Since the borated stainless steel has relatively low heat conductivity, aluminum alloy plates are inserted along with some stainless steel plates to enhance heat removal ability. 2) Cured resin blocks are fitted into the inner shell of the cask for neutron shielding of the cask body. The resin blocks are surrounded by an aluminum casing which transfers heat of stored fuel from the inner shell to the outer shell of the cask. The block type shield structure eliminates the need for welding the heat transfer fins to the inner and outer shells. The weldless structures of the HDP69B lead to its enhanced manufacturability, but they complicate the heat transfer characteristics because there are gaps between such components as the aluminum casing and inner/outer shells. We carried out full-scale prototyping of the HDP69B and ran a heat transfer test using the prototype. The purposes of the heat transfer test were to check the heat removal ability of the HDP69B and to verify the safety analysis model for heat removal. Results of the heat transfer test and optimized analysis model for heat transfer characteristics of the HDP69B are the focus of this paper. The heat transfer test is summarized as follows. Sixty nine heaters simulating the shape and heat power of spent fuel assemblies were inserted into the fuel basket. After replacing the inner atmosphere with 0.1 MPa of helium, the heat transfer test was started. About 7 days were required to equilibrate the temperature distribution. The temperature at the center of the basket was 194 C. The results confirmed the HDP69B had sufficient heat removal ability. The

  8. Full-scale prototyping of the Hitachi dual-purpose metal cask and verification of its heat transfer characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, N.; Ishida, N.; Ootsuka, M.; Kamoshida, M.; Hiranuma, T.; Doumori, S.; Hoshikawa, T.; Shimizu, M.; Kashiwakura, J.; Hayashi, M. [Hitachi, Ltd., Hitachi (Japan)

    2004-07-01

    Hitachi has been developing dual-purpose metal casks for transport and storage of spent nuclear fuels. The Hitachi cask, HDP69B can store 69 BWR fuel assemblies. The cask features are as follows. 1) The fuel basket is assembled mainly with plates of borated stainless steel. The plates are not welded, but cross-inserted into each other like the dividers in an egg carton. Since the borated stainless steel has relatively low heat conductivity, aluminum alloy plates are inserted along with some stainless steel plates to enhance heat removal ability. 2) Cured resin blocks are fitted into the inner shell of the cask for neutron shielding of the cask body. The resin blocks are surrounded by an aluminum casing which transfers heat of stored fuel from the inner shell to the outer shell of the cask. The block type shield structure eliminates the need for welding the heat transfer fins to the inner and outer shells. The weldless structures of the HDP69B lead to its enhanced manufacturability, but they complicate the heat transfer characteristics because there are gaps between such components as the aluminum casing and inner/outer shells. We carried out full-scale prototyping of the HDP69B and ran a heat transfer test using the prototype. The purposes of the heat transfer test were to check the heat removal ability of the HDP69B and to verify the safety analysis model for heat removal. Results of the heat transfer test and optimized analysis model for heat transfer characteristics of the HDP69B are the focus of this paper. The heat transfer test is summarized as follows. Sixty nine heaters simulating the shape and heat power of spent fuel assemblies were inserted into the fuel basket. After replacing the inner atmosphere with 0.1 MPa of helium, the heat transfer test was started. About 7 days were required to equilibrate the temperature distribution. The temperature at the center of the basket was 194 C. The results confirmed the HDP69B had sufficient heat removal ability. The

  9. AREVA NP Inc next generation fresh UO2 fuel assembly shipping cask: SCALE - CRISTAL comparisons lead to safety criticality confidence

    International Nuclear Information System (INIS)

    AREVA NP as a worldwide PWR fuel provider has to have a fleet of fresh UO2 shipping casks being agreed within a lot of countries including USA (France), Germany, Belgium, Sweden, China, and South Africa... and to accommodate foreseen EPR Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector decided to develop an up to date shipping cask gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and local Safety Authorities. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: -Preferential flooding - Fuel rod lattice pitch expansion for full length of fuel assemblies - Neutron absorber penalty -... Well known computer codes, American SCALE package and French CRISTAL package, were used to check configurations reactivity and to ensure that both codes lead to coherent results. Basic spectral calculations are based on similar algorithms with specific microscopic cross sections ENDF/BV for SCALE and JEF2.2 for CRISTAL. The main differences between the two packages is on one hand SCALE's three dimensional fuel assembly geometry is described by a pin by pin model while an homogenized fuel assembly description is used by CRISTAL and on the other hand SCALE is working with either 44 or 238 neutron energy groups while CRISTAL is with a 172 neutron energy groups. Those two computer packages rely on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask with two fuel assemblies is designed to maximize fuel isolation inside a cask and with neighboring ones even for large array configuration cases. Proven industrial products are used: - BoralTM as neutron absorber - High density polyethylene (HDPE) or Nylon as neutron moderator - Foam as thermal and mechanical protection The cask is designed to handle the complete

  10. AREVA NP next generation fresh UO2 fuel assembly shipping cask: SCALE - CRISTAL comparisons lead to safety criticality confidence

    International Nuclear Information System (INIS)

    AREVA NP as a worldwide PWR fuel provider has to have a fleet of fresh UO2 shipping casks being agreed within a lot of countries including USA, France, Germany, Belgium, Sweden, China, and South Africa - and to accommodate foreseen EPR Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector decided to develop an up-to-date shipping cask (so called MAP project) gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and local Safety Authorities. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: - Preferential flooding; - Fuel rod lattice pitch expansion for full length of fuel assemblies; - Neutron absorber penalty; -... Well known computer codes, American SCALE package and French CRISTAL package, were used to check configurations reactivity and to ensure that both codes lead to coherent results. Basic spectral calculations are based on similar algorithms with specific microscopic cross sections ENDF/BV for SCALE and JEF2.2 for CRISTAL. The main differences between the two packages is on one hand SCALE's three dimensional fuel assembly geometry is described by a pin by pin model while an homogenized fuel assembly description is used by CRISTAL and on the other hand SCALE is working with either 44 or 238 neutron energy groups while CRISTAL is with a 172 neutron energy groups. Those two computer packages rely on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask with two fuel assemblies is designed to maximize fuel isolation inside a cask and with neighboring ones even for large array configuration cases. Proven industrial products are used: - BoralTM as neutron absorber; - High density polyethylene (HDPE) or Nylon as neutron moderator; - Foam as thermal and mechanical protection. The cask is

  11. Preliminary safety analysis of criticality for dual-purpose metal cask under dry storage conditions in South Korea

    International Nuclear Information System (INIS)

    Highlights: • DPC is under development led by Korea Radioactive Waste Agency in South Korea. • The results of criticality analysis with respect to design requirements. • The keff under normal and off-normal conditions were 0.36 and 0.46, respectively. • In addition, the keff under a postulated accident condition was evaluated to be 0.94. - Abstract: A dual-purpose metal cask is under development led by Korea Radioactive Waste Agency (KORAD) in Korea, for the dry interim storage and long-distance transportation. This cask comprises a main body made of carbon steel and a stainless steel Dry Shielded Canister (DSC), with stainless steel baskets inside to contain spent fuel assemblies. In this study, nuclear criticality safety analysis was conducted as a part of safety assessment of the metal cask. Analysis to show criticality safety in accordance with regulatory requirements of PWR spent fuel storage was carried out. 10CFR72.124 “Criteria for nuclear criticality safety” and the Regulatory Guide of the American Nuclear Society, ANSI/ANS-57.9 “Design Criteria for an Independent Spent Fuel” and US NRC's “Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility” were employed as regulatory standard and criteria. This paper shows results of criticality analysis with respect to each designated criterion with modeling of a virtual nuclear fuel assembly and a cask body that induces the maximum reactivity among various design basis fuels of the metal cask. In addition, the sensitivity analysis of nuclear criticality taking into account the various modeling deviation such as manufacturing tolerance and modeling assumptions of conventional models was carried out to ensure the reliability of the analysis result. The criticality evaluation result of the metal cask and the maximum keff under normal and off-normal conditions were 0.36884 and 0.46255, respectively. The maximum keff under a postulated accident condition triggering

  12. Analisis dan Desain BI-Dashboard Monitoring Realisasi Daftar Isian Pelaksanaan Anggaran (DIPA pada Kantor Pelayanan Perbendaharaan Negara (KPPN

    Directory of Open Access Journals (Sweden)

    Ernestina Rahmanasari

    2013-09-01

    Full Text Available Transaksi keuangan badan pemerintahan tingkat daerah dipertanggungjawabkan kepada Kementrian Keuangan melalui Kantor Pelayanan Perbendaharaan Negara (KPPNsetiap bulan dalam bentuk Laporan Keuangan Pemerintah Pusat (LKPP. Laporan Keuangan, didukung dengan teknologi Business Intelligence(BI dashboarddapat dimanfaatkan sebagai data-driven decision support system yang mendorong performa kinerja. Analisis dan desainterhadap pembuatan BI dashboard menunjukkanbagaimana Laporan Keuangan dapat digunakan untuk monitoring realisasi penyerapan dana DIPA dan kinerja instansi secara tepat waktu dan interaktif.Analisis terdiri dari analisis Indikator Kinerja Utama (IKU dan analisis Laporan Keuangan. IKUpada analisis dan desain ini mencakup ketercapaian estimasi dengan penerimaan, ketercapaian pagu dengan belanja, ketepatan waktu pengiriman rekonsiliasi oleh instansi, validasi, akurasi, dan ketepatan data. Desain terdiri dari katalog dashboard, alur analisis, sumber data, desain User-Interface, dan desain objek dashboard. Sedangkan dashboard yang dirancang menggunakan aplikasi Vera sebagai Enterprise System, mengakses basis data MySQL, dan menampilkan data dengan teknologi data-driven business-intelligence Qlikview. Desain aplikasi yang telah dibuat akan menampilkan 4 (empat Tab yaitu: Dashboard utama, Analisis, Monitoring, dan Laporan

  13. ANALISIS STEGANOGRAFI METODE LEAST SIGNIFICANT BIT (LSB DENGAN PENYISIPAN SEKUENSIAL DAN ACAK SECARA KUANTITATIF DAN VISUAL

    Directory of Open Access Journals (Sweden)

    Erwin Yudi Hidayat

    2013-08-01

    Full Text Available Penelitian ini bertujuan untuk melakukan analisis terhadap steganografi Least Significant Bit (LSB yang mampu menyisipkan pesan secara sekuensial dan acak. Analisis dilakukan untuk mengetahui penyisipan yang manakah yang memiliki kemampuan paling baik. Secara kuantitatif, Peak Signal to Noise Ratio (PSNR digunakan untuk mengukur kualitas citra. Sedangkan secara visual, steganalisis Enhanced LSB dimanfaatkan untuk mengetahui teknik mana yang mampu menyisipkan pesan tanpa mudah dideteksi. Hasil percobaan menunjukkan, penyisipan secara acak memiliki kemampuan lebih baik daripada penyisipan secara sekuensial. Kata Kunci: steganografi, acak, sekuensial, PSNR, Enhanced LSB

  14. CASKETSS-DYNA2D: a nonlinear impact analysis computer program for nuclear fuel transport casks in two dimensional geometries

    International Nuclear Information System (INIS)

    A nonlinear impact analysis computer program DYNA2D, which was developed by Hallquist, has been introduced from Lawrence Livermore National Laboratory for the purpose of using impact analysis of nuclear fuel transport casks. DYNA2D has been built in CASKETSS code system (CASKETSS means a modular code system for CASK Evaluation code system for Thermal and Structural Safety). Main features of DYNA2D are as follows; (1) This program has been programmed to provide near optimal speed on vector processing computers. (2) An explicit time integration method is used for fast calculation. (3) Many material models are available in the program. (4) A contact-impact algorithm permits gap and sliding along structural interfaces. (5) A rezoner has been embedded in the program. (6) The graphic program for representations of calculation is provided. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  15. The role of sensor directed, model-based control in robotic handling of nuclear waste casks and materials

    International Nuclear Information System (INIS)

    This paper discusses the results from several projects investigating the application of intelligent machine technologies to remote handling of nuclear waste casks and materials. The importance of computer models of the robot, its environment and their interactions is focused upon. Integration of such models into the sensor based control of robot systems results in significant increases in the capabilities of commercial robots by allowing tuning of robot performance to the task

  16. SAVIT: a dymanic model to predict vibratory motion within a spent fuel shipping cask; rail car system

    International Nuclear Information System (INIS)

    A dynamic model of a spent fuel shipping cask-rail car system has been developed to provide estimates of the vibratory motion of LWR spent fuel assemblies during transport and to estimate the effects of this motion on the condition of the assemblies when they arrive at receiving and storage facilities. Results of preliminary test computations are presented to illustrate the capabilities of the model

  17. Analisis komoditas unggulan perikanan budidaya Kabupaten Pidie Jaya

    Directory of Open Access Journals (Sweden)

    Farok Afero

    2015-06-01

    Full Text Available Abstract. Snapper (Lates calcarifer, grouper (Epinephelus coioides, tiger shrimp (Penaeus monodon, vannamei shrimp (Litopenaeus vannamei and tilapia (Oreochromis niloticus are leading commodity worthly cultivated in Pidie Jaya. Bandar Baru and Tringgadeng an appropriate areas for black tiger shrimp while Jangka Buya and Ulim approriate areas for vannamei cultivation. AHP analysis showed black tiger shrimp is top priority based on the economic value while vannamei shrimp is top priority based on enterprise sustainability. Financial analysis of snapper, grouper, black tiger shrimp, vannamei shrimp and tilapia farming generated positive cash flow and NPV, IRR > 100%, the ratio of benefit to cost of production > 1,30 and payback period of investment costs  < 1 year, thus demonstrating the feasibility of cultivation of these leading commodities. Vannamei shrimp cultivation showed positive prospect as long the market offers premium price. Grouper and snapper had a positive outlook because high demand of high-quality fish in the international market. Keywords: Leading commodity; financial analysis; aquaculture; Pidie Jaya Abstrak. Komoditas kakap (Lates calcarifer, kerapu (Epinephelus coioides,udang windu (Penaeus monodon, udang vannamei (Litopenaeus vannamei dan nila (Oreochromis niloticus adalah komoditas unggulan yang layak dibudidaya di Kabupaten Pidie Jaya. Bandar Baru dan Tringgadeng merupakan kawasan yang layak untuk pengembangan komoditas udang windu sedangkan Jangka Buya dan Ulim layak untuk pengembangan komoditas udang vannamei. Analisis AHP menunjukkan komoditas udang windu menjadi prioritas utama untuk dikembangkan berdasarkan nilai ekonomi sedangkan udang vannamei menjadi prioritas utama berdasarkan keberlanjutan usaha. Analisis indikator keuangan budidaya kakap, kerapu, udang windu, vannamei dan nila menunjukkan usaha budidaya komoditas unggulan menghasilkan arus kas kumulatif dan NPV positif, nilai  IRR diatas 100%, rasio keuntungan

  18. LA COMUNICAZIONE POLITICA SUI SOCIAL NETWORK: UN’ANALISI LINGUISTICA

    Directory of Open Access Journals (Sweden)

    Daniele Spoladore

    2014-07-01

    Full Text Available Questo lavoro si propone di analizzare le scelte linguistiche dei soggetti politici che utilizzano Facebook e Twitter per rivolgersi al sempre più frammentato panorama dei potenziali elettori. In particolare, analizza le tendenze linguistiche comuni che questi social network hanno fatto emergere nel rapporto tra politica, web e lettori. Cominciando dall’analisi della presenza di parlamentari e senatori su Facebook e Twitter, si cerca di riassumere le principali caratteristiche delle due piattaforme, per giungere ad una classificazione dei testi prodotti attraverso di esse; si analizza la scrittura di post e tweet, sottolineando la presenza di espedienti tipici del mezzo e valutando il loro effetto sui lettori, e si studiano le scelte sintattiche e morfosintattiche in relazione alla struttura e alle possibilità dei due social network. In ultima analisi si osservano le scelte lessicali dei soggetti politici, studiandone i campi semantici e la quantità di tecnicismi. Infine, si cerca di compiere una valutazione delle due differenti tipologie di trasmesso scritto rinvenute nei campioni, osservando come ognuna di esse abbia uno scopo ben preciso nell’economia della comunicazione politica sui social network. Political communication policy in social networks: a language analysis  Daniele Spoladore This paper aims to analyze the linguistic choices of politicians who use Facebook and Twitter to address the increasingly fragmented landscape of potential voters. In particular, it analyzes the common language trends that these social networks have revealed in the relationship between politics, the web and readers. Starting from the analysis of the presence of MPs and senators on Facebook and Twitter, we try to summarize the key features of the two platforms, in order to arrive at a classification of the texts produced using them.  We analyzed posts and tweets, emphasizing the  typical characteristics of these means and evaluating their effect on

  19. Analisis Kinerja Operasional Kereta Api Sriwedari Ekspress Jurusan Solo - Yogya

    Directory of Open Access Journals (Sweden)

    Bayu Rosida Sumantri

    2014-03-01

    Full Text Available Pada bulan Nopember 2012 PT. Kereta Api Indonesia (Persero DAOP 6 Yogyakarta mengoperasikan Kereta Api Sriwedari Ekspress dengan rute perjalanan Yogya – Solo, hal ini dilakukan untuk mengantisipasi lonjakan penumpang akibat pemangkasan rute perjalanan Kereta Api Prambanan Ekspress dari tiga belas kali perjalanan menjadi enam kali perjalanan. Menurut Humas PT. Kereta Api Indonesia (Persero DAOP 6 Yogyakarta, dibukanya rute baru Kereta Api Sriwedari Ekspress yang melayani delapan kali perjalanan hanya  dengan rute Yogya – Solo saja diharapkan kereta ini dapat memfasilitasi pergerakan penumpang antar dua daerah tersebut. Namun untuk mengetahui kinerja dari kereta api ini sesuai dengan Surat Keterangan Dirjen Perhubungan Darat No. 687 Tahun 2002, perlu ditinjau dari segi faktor muat, jumlah penumpang yang diangkut, waktu tunggu penumpang, ketepatan waktu kedatangan dan keberangkatan kereta dan kenyamanan penumpang. Dalam penelitian digunakan metode survey untuk mendapatkan nilai – nilai dari kinerja kereta api Sriwedari berupa survey observasi atau pengamatan lapangan.  Adapun  analisis yang digunakan untuk perhitungan waktu tempuh, waktu henti dan waktu tunda menggunakan uji hipotesis 1 sample t-test, karena uji ini paling memenuhi untuk melihat diterima atau tidaknya keterlambatan dari waktu kereta. Sedangkan untuk perhitungan load factor dan kenyamanan duduk dan berdiri berdasarkan perhitungan kapasitas dari Vukan R. Vuchic. Dari hasil analisis didapatkan waktu tempuh rata-rata untuk arah Yogyakarta selama 1 jam 14 menit dan untuk arah Solo selama 1 jam 12 menit. Waktu henti yang didapat dari seluruh jadwal perjalanan kereta arah Yogyakarta sebesar 3 menit, untuk arah Solo sebesar 5 menit. Waktu tunda dari seluruh perjalanan kereta Api Sriwedari untuk arah Yogyakarta didapat nilai waktu tunda kedatangan sebesar 22 menit dan nilai waktu tunda keberangkatan sebesar 20 menit, untuk arah Solo didapat nilai waktu tunda kedatangan sebesar 34

  20. Operational assessment of the transnuclear TN-9 truck spent fuel shipping cask: studies and research concerning BNFP

    International Nuclear Information System (INIS)

    This report presents the results of an operational assessment of the Transnuclear Inc., TN-9 spent fuel cask. This packaging system transports seven current generation boiling-water-reactor nuclear fuel assemblies in a truck shipping mode. The studies were performed at the Barnwell Nuclear Fuel Plant by employees of Allied-General Nuclear Services. The work was funded by the Department of Energy during fiscal year 1981. The cooperation of Transnuclear in this effort is gratefully acknowledged. The study is based on repeated simulated unloading runs of TN-9. Specific tasks and areas of study included: (1) sequential dry-run handling operations under simulated unloading conditions, (2) detailed time and manpower studies, (3) estimates of operator radiation exposure, (4) a general evaluation of the cask system capabilities as they relate to unloading and loading facility operations, and (5) preparation of operating procedures for both unloading (confirmed by practice runs) and loading (yet to be confirmed). Also included is general information on the cask, auxiliary equipment, and the Certificate of Compliance

  1. The Application of Fishbone Diagram Analisis to Improve School Quality

    Directory of Open Access Journals (Sweden)

    Slameto Slameto

    2016-06-01

    Full Text Available The research problems are: 1 What steps are to take in a program development aimed at improving the quality of school using a fishbone analysis? 2 Is the program model using fishbone analysis  effective and efficient in meeting the school’s needs to improve its quality? This is research and developmental which comprises 3 phases, namely Preliminary Study, Model Development, and Evaluation/Model Testing. The qualitative data come from the input of management experts and the result of interviews/FGD with stakeholders. The quantitative data are obtained from the assessment of management experts on the product draft, the observation sheets for the field study on the standards of education, and the try out. Data analisis on the validation result uses a descriptive analysis technique. Data from the questionnaire are analyzed by descriptive statistical technique. The results are: 1 the developmental steps in the school quality improvement program by way of fish bone analysis have gone through 6 phases, 2 the research product using fish bone diagram has proved to be simple, applicable, important, controllable, as well as adaptable. Furthermore, it is communicable, so that it has been effective and efficient in meeting the school’s needs for making its educational quality improved.

  2. ANALISIS TINGKAT PENDAPATAN PEDAGANG CANANG DI PASAR BADUNG

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    Surya Dewi Rustariyuni

    2015-12-01

    Full Text Available This studyanalyzes thecharacteristic of “canang”sellers, their economicactivities, their revenues, the factors which influencethem to chooseBadung Market for selling their products and theircontributions in their household economy. As a matter of fact,all of Hindu ismpeoplein Bali use canang everyday for conductingall ceremonies, but only few people whomake canang. It has occured foryears. The method used for analizing the data isdescriptive analysis. The result shows thatthe canang sellersare mostlyfrom Bali. They are dominated by married women atthe age of 15 up to 35 years old. Then, theireducational backgroundaresenior high schools. They start selling early in the morning and stop their activities in the evening.Finally, usually the sellers get much money when there is a big ceremony.Penelitian ini menganalisis karakteristik penjual "canang", kegiatan ekonominya, pendapatannya, faktor-faktor yang mempengaruhi penjual canang untuk memilih Pasar Badung sebagai tempat untuk menjual canangnya dan kontribusi penjual canang dalam perekonomian rumah tangga. Sebenarnya, semua orang Hindu di Bali menggunakan canang setiap hari untuk melakukan semua upacara, tetapi hanya sedikit orang yang membuat canang. Ini telah terjadi selama bertahun-tahun. Metode yang digunakan untuk menganalisis data disini adalah analisis deskriptif. Hasil penelitian menunjukkan bahwa para penjual canang sebagian besar dari Bali. Mereka didominasi oleh perempuan yang sudah menikah danusianya antara 15 hingga 35 tahun. Kemudian, latar belakang pendidikan mereka adalah sekolah menengah atas. Mereka mulai menjual canang di pagi hari dan selesai di malam hari. Biasanya para penjual mendapatkan uang banyak ketika ada upacara besar.

  3. Interface Issues Arising in Interim Storage Facilities Using Storage/Transport Dual Purpose Dry Metal Casks in Japan. Annex VIII

    International Nuclear Information System (INIS)

    The annual amount of spent fuels (SFs) discharged by the operation of commercial reactors nowadays is estimated to be around 10 000 tU level worldwide. While the amount of SFs already reprocessed account about one-third, the rest are currently stored in storage facilities, typically, in wet pools attached to nuclear power plants (NPPs). Cumulative amount of SFs stored is estimated to be about 250 000 tU by 2010 (I. Hanaki, Japan). While wet pool system is dominant in storage facility designs, new design concepts for storage facilities have been continuously developed. One of these new designs is that using dual purpose dry metal casks. “Dual” here means that the casks are not only designed as storage containers, but also designed as transport containers that will satisfy relevant regulatory requirements for transport of radioactive materials such as TS-R-1. Advantage of adopting such “dual” design in storage facilities lies in that this could contribute to reduce the burden associated with handling operations, because, under such designs, SFs once loaded into casks can easily be “transported” to storage facilities, and after storage of several decades, they can again be “transported” to their destinations, regardless they are reprocessing facilities or final disposal sites. Other than these, adopting this kind of design can reduce the amount of radioactive wastes discharged through storage operation, thus can reduce operation costs while maintaining safety level. In Japan, where 53 commercial NPPs are now in operation and with the annual amount of SFs produced sums up to about 1000 tU, keen needs are perceived among SFs producers (namely, utilities) to secure adequate SFs storage capacity. Therefore, a new application for constructing storage facility of 3000 tU scale in Mutsu city, located in northern part of Aomori prefecture, has been submitted in March 2007 by a subsidiary company of utilities named RFS (Recyclable Fuel Storage Company), using

  4. Cinemateca de Vertigo. El delirio de la mirada: Vertigo (1958, Alfred Hitchcock) Analisis de una secuencia

    OpenAIRE

    Castro de Paz, José Luis

    1992-01-01

    Castro De Paz, JL. (1992). Cinemateca de Vertigo. El delirio de la mirada: Vertigo (1958, Alfred Hitchcock) Analisis de una secuencia. Vértigo. Revista de cine. (2):43-48. http://hdl.handle.net/10251/42923. 43 48 2

  5. ANALISIS PENGARUH CORPORATE SOCIAL RESPONSIBILITY TERHADAP CITRA PERUSAHAAN ( STUDI KASUS PT. International Nickel Indonesia Tbk)

    OpenAIRE

    ULVA, -

    2012-01-01

    ULVA. Analisis Pengaruh Corporate Social Responsibility terhadap Citra Perusahaan (Studi Kasus PT. International Nickel Indonesia Tbk), di bawah bimbingan Drs. M. Ishak Amsari, M.Si, Ak dan Drs. Agus Bandang, M.Si, Ak. Kata Kunci : Corporate Social Responsibility, Citra Perusahaan Penelitian ini dilatarbelakangi oleh semakin berkembangnya perusahaan yang juga memunculkan akan kesadaran tanggung jawab sosial perusahaan terhadap lingkungan dan ...

  6. ANALISIS LABELLING PEREMPUAN DENGAN TEORI FEMINISME PSIKOANALISIS: STUDI KASUS MAJALAH REMAJA OLGA!

    OpenAIRE

    - Muashomah

    2013-01-01

    Labelling perempuan dalam majalah remaja merupakan salah satu tindakan media yang merugikan perempuan. Dalam tulisan ini, penulis mengkaji label-label perempuan, bentuk labelling, analisis teori feminisme psikoanalisis terhadap labelling untuk perempuan dalam majalah remaja. Dalam penelitian ini penulis menggunakan metode semiotik dan penelitian dilakukan terhadap majalah Olga. Hasil penelitian menunjukkan bahwa praktek pelabelan terhadap perempuan yang dilakukan oleh majalah remaja ditujukan...

  7. ANALISIS PENGARUH KUALITAS FINANSIAL PERUSAHAAN, KUALITAS AUDITOR DAN KUALITAS PEREKONOMIAN TERHADAP OPINI AUDIT (GOING CONCERN)

    OpenAIRE

    Baqarina Hadori; Bambang Sudibyo

    2014-01-01

    Abstrak: Analisis Pengaruh Kualitas Finansial Perusahaan, Kualitas Auditor, Dan Kualitas Perekonomian Terhadap Opini Audit (Going Concern). Penelitian ini bertujuan untuk menguji pengaruh faktor keuangan dan faktor non-keuangan terhadap pemberian opini audit going concern oleh auditor. Faktor keuangan yang diuji adalah profitabilitas, likuiditas, solvabilitas, pertumbuhan penjualan tahunan, dan pertumbuhan harga saham. Sedangkan faktor non-keuangan yang diuji adalah kualitas auditor dan kuali...

  8. ‘Usi’ e ‘abusi’ nel diritto: Una riflessione critica sulla normativa in materia di analisi genetiche

    Directory of Open Access Journals (Sweden)

    Salardi Silvia

    2013-01-01

    Full Text Available Al centro di questo contributo vi è l’analisi eticogiuridica degli atti normativi disciplinanti le analisi genetiche in vari contesti, ad esempio, ricerca medica, terapia, medicina legale e così dicendo. Lo scopo è di mettere in evidenza i valori ai quali sono state improntate alcune risposte normative. Pertanto, dopo una ricognizione delle varie tipologie di analisi genetiche e dei loro possibili impieghi, il presente lavoro confronta i testi normativi internazionali, europei e nazionali (Austria, Francia, Germania, Svizzera, al fine di individuare la strada percorsa e da percorrere per salvaguardare il più possibile certi valori ritenuti fondamentali per la preservazione sia dell’autonomia individuale, sia dell’eguaglianza tra i consociati. Si concluderà che non tutte le norme che disciplinano le analisi genetiche possono considerarsi rispettose dei diritti fondamentali garantiti a tutti gli individui.

  9. Analisis Faktor-faktor Yang Mempengaruhi Dividend Payout Ratio Pada Industri Barang Konsumsi Di Bursa Efek Indonesia.

    OpenAIRE

    Ketaren, Roby Juahta

    2011-01-01

    Penelitian ini bertujuan untuk menguji dan menganalisis faktor-faktor yang mempengaruhi Dividend Payout Ratio pada industri barang konsumsi di Bursa Efek Indonesia. Pada penelitian ini digunakan analisis fundamental perusahaan melalui analisis rasio keuangan antara lain Cash Position (CP), Debt to Equity Ratio (DER), Return on Assets (ROA), Growth Potential (GP), Firm Size, dan Dividend Payout Ratio (DPR). Periode penelitian dimulai dari tahun 2006-2009. Hipotesis yang dikemukakan adalah Cash...

  10. BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

    International Nuclear Information System (INIS)

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions

  11. THERMLIB: a material property data library for thermal analysis of radioactive material transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    The paper describes an heat conduction data library and graphical program for analysis of radioactive material transport casks. More than 1000 of material data are compiled in the data library which was produced by Lawrence Livermore Laboratory. Thermal data such as, density, thermal conductivity, specific heat, phase-change or solid-state, transition temperature and latent heat have been tabulated. Using this data library, a data library processing program THERMLIB for thermal analysis has been developed. Main features of THERMLIB are as follows: (1) data have been tabulated against temperature, (2) more than 1000 material data are available, (3) it is capable of graphical representations for thermal data and (4) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows) are available for use of THERMLIB. In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides an user`s guide for computer program and input data for THERMLIB. (author)

  12. Large deformation inelastic analysis of impact for shipping casks. [DYNA3D Code

    Energy Technology Data Exchange (ETDEWEB)

    Charman, C.M.; Grenier, R.M. (General Atomic Co., San Diego, CA (USA)); Nickell, R.E. (Applied Science and Technology, Poway, CA (USA))

    1982-09-01

    This paper describes the use of two- and three-dimensional nonlinear finite element computer programs to design a radioative material transportation cask to withstand a drop of 30 feet onto an unyielding surface. Because of recent advancement in the area of non-linear finite element code development, the use of such codes for an iterative design process is becoming practicable. The paper begins with a section dealing with a two-dimensional side drop analysis and is followed by a discussion of the general capabilities of DYNA3D and a brief discussion of the implementation of the code on a computational mainframe unlike any for which the developer had intended. Then, a section on three-dimensional models of center-of-gravity over a corner impact follows, which introduces design features such as bolted closures, internal impact limiter, seals and shear rings. Figs. showing the deformed model grids are included. Stress and strain results are given in the subsequent section. Finally, we interpret these results in terms of possible rules being developed by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code committees.

  13. Acoustic emission detection with fiber optical sensors for dry cask storage health monitoring

    Science.gov (United States)

    Lin, Bin; Bao, Jingjing; Yu, Lingyu; Giurgiutiu, Victor

    2016-04-01

    The increasing number, size, and complexity of nuclear facilities deployed worldwide are increasing the need to maintain readiness and develop innovative sensing materials to monitor important to safety structures (ITS). In the past two decades, an extensive sensor technology development has been used for structural health monitoring (SHM). Technologies for the diagnosis and prognosis of a nuclear system, such as dry cask storage system (DCSS), can improve verification of the health of the structure that can eventually reduce the likelihood of inadvertently failure of a component. Fiber optical sensors have emerged as one of the major SHM technologies developed particularly for temperature and strain measurements. This paper presents the development of optical equipment that is suitable for ultrasonic guided wave detection for active SHM in the MHz range. An experimental study of using fiber Bragg grating (FBG) as acoustic emission (AE) sensors was performed on steel blocks. FBG have the advantage of being durable, lightweight, and easily embeddable into composite structures as well as being immune to electromagnetic interference and optically multiplexed. The temperature effect on the FBG sensors was also studied. A multi-channel FBG system was developed and compared with piezoelectric based AE system. The paper ends with conclusions and suggestions for further work.

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

    International Nuclear Information System (INIS)

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

  15. Stress corrosion cracking of stainless-steel canister for concrete cask storage of spent fuel

    Science.gov (United States)

    Tani, Jun-ichi; Mayuzumi, Masami; Hara, Nobuyoshi

    2008-09-01

    Resistance to external stress corrosion cracking (ESCC) and crevice corrosion were examined for various candidate canister materials in the spent fuel dry storage condition using concrete casks. A constant load ESCC test was conducted on the candidate materials in air after deposition of simulated sea salt particles on the specimen gage section. Highly corrosion resistant stainless steels (SS), S31260 and S31254, did not fail for more than 46 000 h at 353 K with relative humidity of 35%, although the normal stainless steel, S30403 SS failed within 500 h by ESCC. Crevice corrosion potentials of S31260 and S31254 SS became larger than 0.9 V (SCE) in synthetic sea water at temperatures below 298 K, while those of S30403 and S31603 SS were less than 0 V (SCE) at the same temperature range. No rust was found on S31260 and S31254 SS specimens at temperatures below 298 K in the atmospheric corrosion test, which is consistent with the temperature dependency of crevice corrosion potential. From the test result, the critical temperature of atmospheric corrosion was estimated to be 293 K for both S31260 and S31254 SS. Utilizing the ESCC test result and the critical temperature, together with the weather station data and the estimated canister wall temperature, the integrity of canister was assessed from the view point of ESCC.

  16. The application of fracture mechanics to the safety assessment of transport casks for radioactive materials

    Energy Technology Data Exchange (ETDEWEB)

    Zencker, U.; Mueller, K.; Droste, B.; Roedel, R.; Voelzke, H. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

    2004-07-01

    BAM is the German responsible authority for the mechanical and thermal design safety assessment of packages for the transport of radioactive materials. The assessment has to cover the brittle fracture safety proof of package components made of potentially brittle materials. This paper gives a survey of the regulatory and technical requirements for such an assessment according to BAM's new ''Guidelines for the Application of Ductile Cast Iron for Transport and Storage Casks for Radioactive Materials''. Based on these guidelines higher stresses than before can become permissible, but it is necessary to put more effort into the safety assessment procedure. The fundamentals of such a proof with the help of the methods of fracture mechanics are presented. The recommended procedure takes into account the guidelines of the IAEA Advisory Material which are based on the prevention of crack initiation. Examples of BAM's research and safety assessment practices are given. Recommendations for further developments towards package designs with higher acceptable stress levels will be concluded.

  17. THERMLIB: a material property data library for thermal analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    The paper describes an heat conduction data library and graphical program for analysis of radioactive material transport casks. More than 1000 of material data are compiled in the data library which was produced by Lawrence Livermore Laboratory. Thermal data such as, density, thermal conductivity, specific heat, phase-change or solid-state, transition temperature and latent heat have been tabulated. Using this data library, a data library processing program THERMLIB for thermal analysis has been developed. Main features of THERMLIB are as follows: (1) data have been tabulated against temperature, (2) more than 1000 material data are available, (3) it is capable of graphical representations for thermal data and (4) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows) are available for use of THERMLIB. In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides an user's guide for computer program and input data for THERMLIB. (author)

  18. Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G.; Morrow, Charles.

    2013-01-01

    The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

  19. ANALISIS PENGARUH ROA, EPS, FINANCIAL LEVERAGE, PROCEED TERHADAP INITIAL RETURN

    Directory of Open Access Journals (Sweden)

    Andhi Wijayanto

    2010-03-01

    Full Text Available Riset ini bertujuan untuk mengetahui pengaruh ROA, EPS, Financial Leverage dan Proceed terhadap initial return. Initial return diperoleh dengan mengukur perbedaan harga pada hari pertama perdangangan di pasar sekunder dengan harga saat IPO. Penelitian ini menduga bahwa ROA, EPS, Proceed mempunyai pengaruh negatif dengan initial return, disisi lain, Financial Leverage diduga mempunyai pengaruh yang positif terhadap initial return. Data pada penelitian ini terdapat dalam prospectus perusahaan. Sampel diambil dengan menggunakan metode purposive sampling dengan dua kriteria yaitu terdiri dari perusahaan yang IPO selama periode tahun 2000-2006 dan underpriced. Dengan kriteria tersebut, 67 perusahaan dijadikan sebagai sampel. Metode analisis menggunakan regresi berganda. Hasil penelitian ini adalah Earning Per-Share (EPS, dan Proceed mempunyai pengaruh negatif dan signifikan terhadap initial return, sedangkan Return on Assets Ratio (ROA, dan Financial Leverage tidak berpengaruh signifikan terhadap initial return. This research aimed to examine the influence of ROA, EPS, Financial Leverage, and Proceed on initial return. Initial return was measured by the difference between the firm’s stock price on the first day in the secondary market and it’s IPO. This research expected that return on assets ratio (ROA, earning per-share (EPS, and proceed negatively associated with initial return. On other hand, financial leverage ratio expected to positively associate with initial return. Data in this study were obtained from company prospectus, ICMD. Sample had been taken by using purposive sampling method with two criterions such as conducted IPO during period 2000-2006 and underpriced. With criterions, 67 companies obtains as sample. The analytical methods used multiple regressions, the empirical result of this research indicate that EPS, and proceed significantly associated with initial returns. Whereas ROA, and financial leverage ratio not

  20. ANALISIS HUBUNGAN JANGKA PANJANG ANTARA ANGGOTA DENGAN KOPERASI JASA KEUANGAN

    Directory of Open Access Journals (Sweden)

    Eko Nur Udin Aziz

    2012-09-01

    Full Text Available Penelitian ini, dilakukan untuk mendeskripsikan dan menganalisis pengaruh Kemampuan Tenaga Pemasaran, Reputasi Lembaga dan Kepuasan Anggota terhadap Hubungan Jangka Panjang yang dimediasi oleh Kepercayaan Anggota, serta pengaruh Kemampuan Tenaga Pemasaran Terhadap Hubungan Jangka Panjang dan pengaruh Kepuasan Anggota terhadap Hubungan Jangka Panjang antara anggota dengan KJKS BMT Bina Ummat Sejahtera. Dari hasil analisis data atas model yang dikembangkan didapatkan nilai indeks pengukuran RMSEA (0.065, GFI (0.928, AGFI (0.822, TLI (0.936 dan CFI (0.968 berada dalam rentang nilai yang baik. Maka model ini dapat diterima karena secara umum model memiliki tingkat goodness of fit yang dapat diterima meskipun nilai chi square dan probabilitas diterima secara marginal. Dari pengujian terhadap enam hipotesis yang diajukan dalam penelitian ini semua hipotesis alternatif didukung yaitu H1, H2, H3, H4, H5 dan H6 dapat didukung.This study aims to describe and to analyze the influence of salesforce ability, Institute reputation and Member Satisfaction toward Long Term Relationships, which are mediated by the trust of the members, as well as the influence of salesforce ability toward Long-Term Relationship and the influence of the members satisfaction toward Long-Term Relationships among members of the BMT KJKS Bina Ummat Sejahtera. Analysis of the data obtained the model developed index measuring the value of RMSEA (0065, GFI (0928, AGFI (0822, TLI (0936 and CFI (0968 is in the range of good value. In addition, this model can be generally accepted as the model which has level of goodness of fit and can be accepted even though the value of chi square and probability is marginally accepted. Examination of the six hypotheses proposed in this study whereas all received as the alternative hypothesis H1, H2, H3, H4, H5 and H6 can be supported.

  1. ANALISIS EFEKTIVITAS PEMBERIAN PINJAMAN PROGRAM PEMBIAYAAN UMKM OLEH KOPERASI

    Directory of Open Access Journals (Sweden)

    Hadi Ismanto

    2014-10-01

    Full Text Available Abstrak: Analisis Efektivitas Pemberian Pinjaman Program Pembiayaan UMKM Oleh Koperasi. Program Pembiayaan memiliki peran yang penting bagi UMKM, namun sering menghadapi masalah penunggakan dan kemacetan pembayaran angsuran. Penelitian ini berupaya untuk mengetahui sebab-sebab tidak lancarnya pengembalian Program Pembiayaan UJKS sehingga diharapkan dapat menyusun strategi yang lebih baik lagi dalam menyeleksi calon peminjam agar angka pinjaman bermasalah dapat ditekan. Populasi dalam penelitian ini adalah semua UMKM yang menjadi Peminjam (peminjam program pembiayaan UJKS Mitra Usaha dan masih tergolong aktif hingga bulan November 2013 dan telah memperoleh fasilitas pembiayaan sekurang-kurangnya enam bulan berjalan. Penelitian ini menemukan bahwa faktor-faktor yang berpengaruh positif terhadap tingkat pengembalian pinjaman (lancar atau menunggak adalah omzet usaha, lama usaha dan nilai plafon pinjaman. Hal ini dapat dijadikan bahan pertimbangan bagi UJKS dalam menjalankan program pembiayaan sehingga menjadi lebih efektif dalam menjalankannya.   Kata Kunci: UMKM, UJKS, Program Pembiayaan, Lancar, dan Macet.   Abstract: Effectiveness Analysis of SMEs Financing Program by Cooperative. Financing programs have an important role to SMEs, but often face the problem of repayment failures and installment payments. This study examines the causes of saving and loan cooperatives’ repayment problems. The population of this study is all SMEs borrowing from Mitra Usaha saving and loan cooperative. Furthermore these SMEs must be categorized as active SMEs until November 2013 and they had been receiving loan for at least six months. The study found that the variables such as: the length of the business, the business volume, and the value of the loan have a positive impact on repayment rate.   Keywords: SMEs, Saving and Loan Cooperative, Financing Program

  2. ANALISIS KEPEMILIKAN JIWA KEWIRAUSAHAAN: EVALUASI OUTCOME PENDIDIKAN MENENGAH DI JAWA

    Directory of Open Access Journals (Sweden)

    Wahyu Purhantara

    2013-10-01

    Full Text Available Abstract: Analisis Kepemilikan Jiwa Kewirausahaan: Evaluasi Outcome Pendidikan Menengah di Jawa. Pendidikan memiliki peran sentral dalam membentuk karakter dan jiwa kewirausahaan. Karena Indonesia memiliki kurang dari 2% dari penduduknya sebagai pengusaha, pendidikan kewirausahaan menjadi hal yang penting. Menurut Drucker, sebuah negara akan makmur jika penduduknya memiliki jiwa kewirausahaan. Pertanyaannya adalah, mengapa semangat kewirausahaan memainkan peran penting dalam pengembangan organisasi, bisnis, dan pengembangan diri? Jawabannya adalah bahwa kewirausahaan melatih orang untuk menjadi mandiri, kreatif, inovatif, kompetitif, berorientasi hasil, menyukai tantangan, bekerja keras, dan sebagainya. Hasil evaluasi terhadap jiwa kewirausahaan berdasarkan hasil proses pendidikan tingkat SMA/SMK di 5 kota di Jawa pada tahun 2011 menunjukkan bahwa mereka tidak memiliki jiwa kewirausahaan seperti yang diharapkan oleh standar kompetensi bagi lulusan SMA / SMK. Keywords: Jiwa Kewirausahaan, Pendidikan Kewirausahaan   Abstract: Entrepreneurial Spirit Analysis: Outcome Evaluation of Secondary Education in Java. Education has central role in forming character and entrepreneurial spirit. Since Indonesia has less than 2% of its inhabitants as entrepreneurs, education of entrepreneurship becomes urgent. According to Drucker, a country would be prosperous if its inhabitants have entrepreneurial spirit. The question is, why entrepreneurial spirit plays an important role in organization development, business, and self development? The answer is that entrepreneurship trains people to become self-supporting, creative, innovative, competitive, result oriented, fond of challenges, hard working, and so on. Result of evaluation on entrepreneurial spirit of high school education outcome in 5 towns in Java in 2011 indicates that they do not have entrepreneurship spirit as expected by the standard of competence for SMA/ SMK graduates. Keywords: entrepreneurial

  3. ANALISIS FAKTOR-FAKTOR YANG MEMPENGARUHI PENERIMAAN PAJAK

    Directory of Open Access Journals (Sweden)

    Phany Ineke Putri

    2013-09-01

    Full Text Available The purpose of this study was to analyze the influence of population, income per capita, inflation and the number of Trade Permit toward billboard tax revenue in Purbalingga. The results of this study is expected to provide benefit and input to the Local Government of Purbalingga, especially Departement of management of financial income and regional asset. Analysis tool used is multiple linear regression (Multiple Linear Regression Method by the method of least squares Ordinary Least Square (OLS. Jointly test results showed that overall independent variables (population, income per capita, inflation, and the number of trade permit together to show their effects on the billboard tax revenue. Adjusted R-Squared value of 0,951, which means 95,1 percent of the four independent variables. While the remaining 4,9 percent is explained by other causes outside the model. The results showed that the per capita income variable and the number of trade permit have a significant positive effect, the inflation variable has a significant negative effect, whereas a number of population variable have a positive effect which is not significant at α = 5 percent of tax revenue in Purbalingga billboard. Tujuan dari penelitian ini adalah untuk menganalisis pengaruh jumlah penduduk, pendapatan per kapita, inflasi dan jumlah surat ijin usaha perdagangan (SIUP terhadap penerimaan pajak reklame di Kabupaten Purbalingga. Manfaat penelitian ini diharapkan menjadi bahan masukan Pemerintah Kabupaten Purbalingga khususnya Dinas Pengelolaan, Pendapatan, Keuangan dan Aset Daerah. Teknik analisis yang digunakan adalah regresi berganda dengan metode Ordinary Least Square (OLS. Hasil penelitian menunjukkan ada pengaruh secara bersama-sama variabel jumlah penduduk, pendapatan per kapita, inflasi dan jumlah surat ijin usaha perdagangan terhadap penerimaan pajak reklame. Koefisien determinasi 95,1% dapat dijelaskan oleh variabel independen terhadap variabel dependen, dan sisanya

  4. Calculative activation analysis of the transport rack for CASTOR {sup registered} casks; Berechnung der Aktivierung eines Transportgestells fuer CASTOR {sup registered} -Behaelter

    Energy Technology Data Exchange (ETDEWEB)

    Tittelbach, S. [Wissenschaftlich-Technische Ingenieurberatung GmbH (WTI), Juelich (Germany); Biedermann, R. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Schmidt-Wohlfarth, Y.; Louia, A. [EnBW Kernkraft GmbH, Philippsburg (Germany)

    2011-07-01

    The transport rack for the internal transport of loaded CASTOR {sup registered} casks before the storage in the intermediate storage facility at the site of the NPP Philippsburg is exposed to neutron irradiation from the cask inventory. Using the Monte Carlo code MCNP the activation rates of the transport rack materials are calculated for typical storage times of the casks in the rack. The long-term activation was also calculated for the continuous use of the transport rack over 10 years. Further topics were the dose rate in the near surrounding of the transport rack after long-term activation and finally the disposability of rack components according to the legal regulations. The maximum contact dose rate was calculated to be below 1 micro Sv/h after 10 years of application. The transport rack can be disposed with large safety margins to the radiation protection limits.

  5. CAPSIZE: A personal computer program and cross-section library for determining the shielding requirements, size, and capacity of shipping casks subject to various proposed objectives

    International Nuclear Information System (INIS)

    A new interactive program called CAPSIZE has been written for the IBM-PC to rapidly determine the likely impact that proposed design objectives might have on the size and capacity of spent fuel shipping casks designed to meet those objectives. Given the burnup of the spent fuel, its cooling time, the thickness of the internal basket walls, the desired external dose rate, and the nominal weight limit of the loaded cask, the CAPSIZE program will determine the maximum number of PWR fuel assemblies that may be shipped in a lead-, steel-, or uranium-shielded cask meeting those objectives. The necessary neutron and gamma shield thicknesses are determined by the program in such a way as to meet the specified external dose rate while simultaneously minimizing the overall weight of the loaded cask. The one-group cross-section library used in the CAPSIZE program has been distilled from the intermediate results of several hundred 1-D multigroaup discrete ordinates calculations for different types of casks. Neutron and gamma source terms, as well as the decay heat terms, are based on ORIGEN-S analyses of PWR fuel assemblies having exposures of 10, 20, 30, 40, 50, and 60 gigawatt days per metric tonne of initial heavy metal (GWD/MTIHM). In each case, values have been tabulated at 17 different decay times between 120 days and 25 years. Other features of the CAPSIZE program include a steady-state heat transfer calculation which will minimize the size and weight of external cooling fins, if and when such fins are required. Comparisons with previously reported results show that the CAPSIZE program can generally estimate the necessary neutron and gamma shield thicknesses to within 0.16 in. and 0.08 in., respectively. The corresponding cask weights have generally been found to be within 1000 lbs of previously reported results. 13 refs., 20 figs., 54 tabs

  6. CAPSIZE: A personal computer program and cross-section library for determining the shielding requirements, size, and capacity of shipping casks subject to various proposed objectives

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    1987-05-01

    A new interactive program called CAPSIZE has been written for the IBM-PC to rapidly determine the likely impact that proposed design objectives might have on the size and capacity of spent fuel shipping casks designed to meet those objectives. Given the burnup of the spent fuel, its cooling time, the thickness of the internal basket walls, the desired external dose rate, and the nominal weight limit of the loaded cask, the CAPSIZE program will determine the maximum number of PWR fuel assemblies that may be shipped in a lead-, steel-, or uranium-shielded cask meeting those objectives. The necessary neutron and gamma shield thicknesses are determined by the program in such a way as to meet the specified external dose rate while simultaneously minimizing the overall weight of the loaded cask. The one-group cross-section library used in the CAPSIZE program has been distilled from the intermediate results of several hundred 1-D multigroaup discrete ordinates calculations for different types of casks. Neutron and gamma source terms, as well as the decay heat terms, are based on ORIGEN-S analyses of PWR fuel assemblies having exposures of 10, 20, 30, 40, 50, and 60 gigawatt days per metric tonne of initial heavy metal (GWD/MTIHM). In each case, values have been tabulated at 17 different decay times between 120 days and 25 years. Other features of the CAPSIZE program include a steady-state heat transfer calculation which will minimize the size and weight of external cooling fins, if and when such fins are required. Comparisons with previously reported results show that the CAPSIZE program can generally estimate the necessary neutron and gamma shield thicknesses to within 0.16 in. and 0.08 in., respectively. The corresponding cask weights have generally been found to be within 1000 lbs of previously reported results. 13 refs., 20 figs., 54 tabs.

  7. ANALISIS FAKTOR PEMANFAATAN POLINDES MENURUT KONSEP MODEL PERILAKU KESEHATAN ”ANDERSON” (Analisis Lanjut Data RISKESDAS 2007

    Directory of Open Access Journals (Sweden)

    Astridya Paramita

    2013-11-01

    dari kajian analisis lanjut data sekunder Riset Kesehatan Dasar (Riskesdas 2007. Tehnik analisis yang digunakan adalah uji chi square dan regresi logistik multiple untuk memperoleh gambaran hubungan antara karakteristik, status sosial rumah tangga, dan kemudahan akses Polindes terhadap pemanfaatan Polindes. Hasil kajian menunjukkan pemanfaaatan Polindes oleh rumah tangga di Indonesia masih rendah dengan alasan tidak butuh pelayanan Polindes. Uji Chi Square menunjukkan terdapat hubungan bermakna antara jarak tempuh, waktu tempuh, klasifikasi desa, pengeluaran per kapita, pekerjaan, pendidikan, dan umur kepala rumah tangga terhadap pemanfaatan Polindes. Hasil uji regresi logistik multiple menunjukkan adanya kecenderungan pemanfaatan Polindes oleh rumah tangga berdasarkan jarak tempuh, waktu tempuh, klasifikasi desa, pengeluaran per kapita, pekerjaan, pendidikan dan umur kepala rumah tangga. Kesimpulan yang dapat diambil dari tulisan ini adalah Polindes lebih tepat didesain untuk kelompok rumah tangga pra sejahtera, berpendidikan rendah, bermata pencaharian petani atau nelayan, dan berlokasi di pedesaan atau daerah terpencil dengan keterbatasan alat transportasi. Kata kunci: karakteristik dan status sosial rumah tangga, aksesibilitas Polindes, pemanfaatan Polindes

  8. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    International Nuclear Information System (INIS)

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications

  9. First burnup credit application including actinides and fission products for transport and storage cask by using French experiments

    International Nuclear Information System (INIS)

    The burnup credit (BUC) methodology for a transport and storage cask application, including actinides and fission products, is implemented at AREVA TN using the French BUC calculation route for pressurized water reactor (PWR) UO2 used fuel. The methodology is based on the connection of the French depletion code DARWIN2 and the French criticality safety package CRISTAL V1. The BUC methodology includes the experimental validation of the computation codes dedicated to the calculation of the used fuel inventory calculations. Indeed, the results of the comparison calculation–experiment (C-E)/E allow to determine either a set of isotopic correction factors (ICFs) for the BUC nuclides considered in the criticality calculation or keff-penalty terms directly used for the definition of the keff-acceptance criterion for the criticality assessment of the transport and storage cask. These ICFs or keff-penalty terms are one of the key of the BUC method to guarantee the conservativeness of the fuel reactivity in safety-criticality calculations using BUC approach. A French BUC program has been developed at CEA/Cadarache in the framework of the CEA-AREVA collaboration in order to validate fuel inventory calculations. This program involves two kinds of experiments: chemical analyses and microprobe measurements of PWR irradiated fuel pins (French PIE program) on one hand, and reactivity worth measurements of the BUC nuclides in the MINERVE reactor on the other hand. This paper highlights, through a first industrial AREVA TN's application of the BUC method, including fission products, that the French PIE program and reactivity worth measurements in MINERVE reactor are suitable for the implementation of BUC in transport and storage cask applications loaded with PWR UO2 used fuels assemblies. (author)

  10. Analysis, scale modeling, and full-scale test of a railcar and spent-nuclear-fuel shipping cask in a high-velocity impact against a rigid barrier

    Energy Technology Data Exchange (ETDEWEB)

    Huerta, M.

    1981-06-01

    This report describes the mathematical analysis, the physical scale modeling, and a full-scale crash test of a railcar spent-nuclear-fuel shipping system. The mathematical analysis utilized a lumped-parameter model to predict the structural response of the railcar and the shipping cask. The physical scale modeling analysis consisted of two crash tests that used 1/8-scale models to assess railcar and shipping cask damage. The full-scale crash test, conducted with retired railcar equipment, was carefully monitored with onboard instrumentation and high-speed photography. Results of the mathematical and scale modeling analyses are compared with the full-scale test. 29 figures.

  11. Analysis, scale modeling, and full-scale test of a railcar and spent-nuclear-fuel shipping cask in a high-velocity impact against a rigid barrier

    International Nuclear Information System (INIS)

    This report describes the mathematical analysis, the physical scale modeling, and a full-scale crash test of a railcar spent-nuclear-fuel shipping system. The mathematical analysis utilized a lumped-parameter model to predict the structural response of the railcar and the shipping cask. The physical scale modeling analysis consisted of two crash tests that used 1/8-scale models to assess railcar and shipping cask damage. The full-scale crash test, conducted with retired railcar equipment, was carefully monitored with onboard instrumentation and high-speed photography. Results of the mathematical and scale modeling analyses are compared with the full-scale test. 29 figures

  12. Validation of CFD-methods to predict heat transfer and temperatures during the transport and storage of casks under a cover

    Energy Technology Data Exchange (ETDEWEB)

    Leber, A. [WTI Wissenschaftlich-Technische-Ingenieurberatung GmbH (Germany); Graf, W. [GNS Gesellschaft fuer Nuklear-Service mbH (Germany); Hueggenberg, R. [GNB Gesellschaft fuer Nuklear-Behaelter mbH (Germany)

    2004-07-01

    With respect to the transport of casks for radioactive material, the proof of the safe heat removal can be accomplished by validated calculation methods. The boundary conditions for thermal tests for type B packages are specified in the ADR based on the regulations defined by the International Atomic Energy Agency. The varying boundary conditions under transport or storage conditions are based on the varying thermal conditions true for different cask types. In most cases the cask will be transported in lying position under a cover (e.g. canopy or tarpaulin) and stored in standing position in an array with other casks. The main heat transport mechanisms are natural convection and thermal radiation. The cover or the storage building are furnished with vents that create an air flow, which will improve the natural convection. Depending on the thermal boundary conditions, the cask design and the heat power, about 50 - 95% of the heat power will be removed from the finned cask surface by natural convection. Consequently the convection by air flow is the main heat transport mechanism. The air flow can be approximated with analytical methods by solving the integral heat and flow balances for the domain. In a stationary state the overpressure due the buoyancy and the pressure loss in the flow resistances are equal. Based on the air flow, the relevant temperatures of the cask can be calculated in an iterative process. Due to the fast development of numerical calculation methods and computer hardware, the use of Computational- Fluid-Dynamics(CFD) calculations plays an important role. CFD-calculations are based on solving the equations of conservation (Navier-Stokes equations) using a finite element mesh or a finite volume mesh of the model. For a finned cask lying under a cover, where the main contributing element for heat removal is natural convection in combination with the thermal radiation, a CFD-calculation can be the most appropriate method. Common CFD-Codes are FLUENT

  13. Validation of CFD-methods to predict heat transfer and temperatures during the transport and storage of casks under a cover

    International Nuclear Information System (INIS)

    With respect to the transport of casks for radioactive material, the proof of the safe heat removal can be accomplished by validated calculation methods. The boundary conditions for thermal tests for type B packages are specified in the ADR based on the regulations defined by the International Atomic Energy Agency. The varying boundary conditions under transport or storage conditions are based on the varying thermal conditions true for different cask types. In most cases the cask will be transported in lying position under a cover (e.g. canopy or tarpaulin) and stored in standing position in an array with other casks. The main heat transport mechanisms are natural convection and thermal radiation. The cover or the storage building are furnished with vents that create an air flow, which will improve the natural convection. Depending on the thermal boundary conditions, the cask design and the heat power, about 50 - 95% of the heat power will be removed from the finned cask surface by natural convection. Consequently the convection by air flow is the main heat transport mechanism. The air flow can be approximated with analytical methods by solving the integral heat and flow balances for the domain. In a stationary state the overpressure due the buoyancy and the pressure loss in the flow resistances are equal. Based on the air flow, the relevant temperatures of the cask can be calculated in an iterative process. Due to the fast development of numerical calculation methods and computer hardware, the use of Computational- Fluid-Dynamics(CFD) calculations plays an important role. CFD-calculations are based on solving the equations of conservation (Navier-Stokes equations) using a finite element mesh or a finite volume mesh of the model. For a finned cask lying under a cover, where the main contributing element for heat removal is natural convection in combination with the thermal radiation, a CFD-calculation can be the most appropriate method. Common CFD-Codes are FLUENT

  14. Creation of a Geant4 Muon Tomography Package for Imaging of Nuclear Fuel in Dry Cask Storage

    Energy Technology Data Exchange (ETDEWEB)

    Tsoukalas, Lefteri H. [Purdue Univ., West Lafayette, IN (United States)

    2016-03-01

    This is the final report of the NEUP project “Creation of a Geant4 Muon Tomography Package for Imaging of Nuclear Fuel in Dry Cask Storage”, DE-NE0000695. The project started on December 1, 2013 and this report covers the period December 1, 2013 through November 30, 2015. The project was successfully completed and this report provides an overview of the main achievements, results and findings throughout the duration of the project. Additional details can be found in the main body of this report and on the individual Quarterly Reports and associated Deliverables of the project, uploaded in PICS-NE.

  15. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    Energy Technology Data Exchange (ETDEWEB)

    Mourao, Rogerio P.; Leite da Silva, Luiz [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil); Miranda, Carlos A.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil); Quintana, Jose F.A.; Saliba, Roberto O. [Comision Nacional de Energia Atomica, Bariloche (Argentina); Novara, Oscar E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  16. Nonlinear dynamic impact analysis for installing a dry storage canister into a vertical concrete cask

    International Nuclear Information System (INIS)

    In this paper, a series of dynamic impact analysis for installing a dry storage canister into a vertical concrete cask (VCC) is performed. The dry storage system considered herein is called HCDSS-69, recently developed by INER and being capable of accommodating 69 bundles of BWR spent nuclear fuels. The impact accident is stemming from a conservative consideration of accidental movement when the canister is being hoisted into a VCC. According to NUREG-0554, the accidental movement is conservatively simulated by 80 mm- and 160 mm-height free-drop motions and then with straight and 2°-oblique impact to a pedestal in VCC. A symmetric fully 3-D finite element model is built and analyzed using the explicit finite element code, LS-DYNA. Geometrical, contact, and material nonlinearities are all taken into account. The analysis result concludes that the permanent deformations of the canister are not severe to affect fuel retrieve after the impact accident and the maximum stress intensity in the canister shell can meet the ASME code appendix F F-1340, preventing the leakage of radioactive materials. The study also found that with properly reducing the wall thickness of the pedestal cylinder, the maximum acceleration and permanent deformation of the canister can be much alleviated, even though the drop height is increased to the double of the required brake distance specified in NUREG-0554. The damages of the pedestal in each analysis are moderate so that the heat transfer condition after the impact accident can be bounded by the off-normal event for half-blockage of air inlets

  17. Developing a structural health monitoring system for nuclear dry cask storage canister

    Science.gov (United States)

    Sun, Xiaoyi; Lin, Bin; Bao, Jingjing; Giurgiutiu, Victor; Knight, Travis; Lam, Poh-Sang; Yu, Lingyu

    2015-03-01

    Interim storage of spent nuclear fuel from reactor sites has gained additional importance and urgency for resolving waste-management-related technical issues. In total, there are over 1482 dry cask storage system (DCSS) in use at US plants, storing 57,807 fuel assemblies. Nondestructive material condition monitoring is in urgent need and must be integrated into the fuel cycle to quantify the "state of health", and more importantly, to guarantee the safe operation of radioactive waste storage systems (RWSS) during their extended usage period. A state-of-the-art nuclear structural health monitoring (N-SHM) system based on in-situ sensing technologies that monitor material degradation and aging for nuclear spent fuel DCSS and similar structures is being developed. The N-SHM technology uses permanently installed low-profile piezoelectric wafer sensors to perform long-term health monitoring by strategically using a combined impedance (EMIS), acoustic emission (AE), and guided ultrasonic wave (GUW) approach, called "multimode sensing", which is conducted by the same network of installed sensors activated in a variety of ways. The system will detect AE events resulting from crack (case for study in this project) and evaluate the damage evolution; when significant AE is detected, the sensor network will switch to the GUW mode to perform damage localization, and quantification as well as probe "hot spots" that are prone to damage for material degradation evaluation using EMIS approach. The N-SHM is expected to eventually provide a systematic methodology for assessing and monitoring nuclear waste storage systems without incurring human radiation exposure.

  18. PENGARUH KESELAMATAN KERJA DAN KESEHATAN KERJA TERHADAP PRODUKTIVITAS KERJA KARYAWAN

    Directory of Open Access Journals (Sweden)

    Muhammad Busyairi

    2014-12-01

    Full Text Available Human Resources (HR is the most important asset in a company. Employees can be a good potential if managed properly and correctly, but employees can also be a burden if the company can not manage it properly. As the increase in the industry, particularly coal mining industry in East Kalimantan is always the problem that arises is the possibility of the occurrence of occupational accidents, occupational diseases, and environmental pollution. The occurrence of occupational accidents, occupational diseases, and environmental pollution due to poor human resource management can reduce employee productivity and company. Research objectives to be achieved is describing the safety and health of employees produkitivitas particularly in the production of PT. XYZ, Loa Kulu aquatic mammal, and to know the significant influence of each variable. Survey respondents were employees in the production of  PT. XYZ, Loa Kulu aquatic mammal with a high of 165 employees. By using the Slovin formula can be determined that a sample of respondents is as many as 99 employees.Based on the description of each item answer the answer of respondents indicated that employees responded positively to the questions given. Multiple regression analysis showed that there were significant effects of occupational safety and health of employee productivity, it can be seen from the significance of each variable is significant <0.05. Significance X1 to Y by 0.000, and the significance of X2 to Y by 0.017. Can be inferred by looking at the results of studies showing that the implementation of occupational safety and health on the productivity of employees of PT. XZY  Loa Kulu Kukar, generally not very good, is expected to be further enhanced supervision.

  19. PENGARUH KESELAMATAN KERJA DAN KESEHATAN KERJA TERHADAP PRODUKTIVITAS KERJA KARYAWAN

    Directory of Open Access Journals (Sweden)

    Muhammad Busyairi

    2014-12-01

    Full Text Available Human Resources (HR is the most important asset in a company. Employees can be a good potential if managed properly and correctly, but employees can also be a burden if the company can not manage it properly. As the increase in the industry, particularly coal mining industry in East Kalimantan is always the problem that arises is the possibility of the occurrence of occupational accidents, occupational diseases, and environmental pollution. The occurrence of occupational accidents, occupational diseases, and environmental pollution due to poor human resource management can reduce employee productivity and company. Research objectives to be achieved is describing the safety and health of employees produkitivitas particularly in the production of PT. XYZ, Loa Kulu aquatic mammal, and to know the significant influence of each variable. Survey respondents were employees in the production of PT. XYZ, Loa Kulu aquatic mammal with a high of 165 employees. By using the Slovin formula can be determined that a sample of respondents is as many as 99 employees.Based on the description of each item answer the answer of respondents indicated that employees responded positively to the questions given. Multiple regression analysis showed that there were significant effects of occupational safety and health of employee productivity, it can be seen from the significance of each variable is significant <0.05. Significance X1 to Y by 0.000, and the significance of X2 to Y by 0.017. Can be inferred by looking at the results of studies showing that the implementation of occupational safety and health on the productivity of employees of PT. XZY Loa Kulu Kukar, generally not very good, is expected to be further enhanced supervision.

  20. Analisis Hubungan Receivable Turnover Ratio, Inventory Turnover Ratio, dan Total Asset Turnover Ratio Dengan Kemampulabaan Perusahaan Pada PTPN III (Persero) Medan

    OpenAIRE

    Panjaitan, Fitry Bertha H.

    2010-01-01

    Tujuan penelitian adalah untuk menganalisis hubungan Receivable Turnover Ratio, Inventory Turnover Ratio, dan Total Asset Turnover Ratio dengan kemampulabaan perusahaan pada PTPN III (Persero) Medan. Penulis menarik hipotesis bahwa Receivable Turnover Ratio, Inventory Turnover Ratio, dan Total Asset Turnover Ratio berhubungan positif dengan kemampulabaan perusahaan pada PTPN III (Persero) Medan. Metode penelitian yang digunakan adalah metode analisis deskriptif dan metode analisis K...

  1. Analisi dei processi di un ambulatorio per i viaggiatori internazionali

    Directory of Open Access Journals (Sweden)

    M.L. Demarchi

    2003-05-01

    materiale informativo anche in lingua straniera. L’analisi ha inoltre rilevato l’assenza di un sistema informativo di feed-back che consentisse il monitoraggio della morbosità delle patologie contratte all’estero, elemento utile in un’ottica di programmazione.

    Conclusioni: l’ABC, grazie all’individuazione delle criticità, ha
    consentito di definire un piano per la riorganizzazione dell’ambulatorio, volto all’ottimizzazione delle risorse disponibili garantendo l’erogazione di prestazioni efficaci ed efficienti.

  2. ANALISIS BANTUAN KREDIT TERHADAP PERKEMBANGAN KELOMPOK USAHA BERSAMA

    Directory of Open Access Journals (Sweden)

    Priyo Harsono

    2012-09-01

    Full Text Available The object of research is KUB Rukun Mina Barokah in Juwana, Pati. The purpose of research is to investigate the differences of KUB Rukun Mina Barokah development before and after getting the credit assistance from the Department of Marine and Fisheries, Pati in terms of capital, labor, thenumber of buyers, the total sales, and profits. The population for this research is 45 small micro enterprises. The type of data collected are primary and secondary data. The methods of data analysis are validity test, reliability test, and Wilcoxon sign rank test statistics. Based on the results of Wilcoxon sign rank test statistics which focuses on the variables of capital, labor, number of buyers, total sales and profit , it is obtained that p-value is 0.000 (0.000 <0,05. It shows that the credit assistance of the Department of Marine and Fisheries, Pati has given the changes in the joint venture group, Rukun Mina Barokah. The changes here means an increase in the variables before and after getting the credit assistance. Therefore, besides financial aid in the form of business loans, collaboration expansion is also needed to develop the business groups (KUB Objek dalam penelitian ini adalah KUB Rukun Mina Barokah di Kecamatan Juwana Kabupaten Pati. Tujuan penelitian ini untuk mengetahui perbedaan perkembangan KUB Rukun Mina Barokah sebelum dan sesudah mendapatkan bantuan kredit dari Dinas Kelautan dan Perikanan Kabupaten Pati ditinjau dari modal usaha, tenaga kerja, jumlah pembeli, total penjualan, dan keuntungan. Populasi penelitian sebanyak 45 usaha mikro kecil.Jenis datanya adalah data primer dan data sekunder.Metode analisis data meliputi uji validitas, uji reabilitas, dan uji statistic pangkat tanda wilcoxon.Berdasarkan hasil uji statistik pangkat tanda wilcoxon, baik variabel modal usaha, variabel tenaga kerja,variabel jumlah pembeli, variabel total penjualan maupun variabel keuntungan didapatkan nilai -p sebesar 0,000 (0,000 < 0,05. Hal itu

  3. ANALISIS METODE KARMARKAR UNTUK MENYELESAIKAN MASALAH PROGRAM LINIER

    Directory of Open Access Journals (Sweden)

    DR Indriani

    2014-06-01

    Full Text Available Penelitian ini bertujuan mengetahui dasar matematis dalam metode Karmarkar, mengetahui penyelesaian masalah program linier dengan metode Karmarkar, serta menganalisis penyelesaian masalah program linier dengan metode simpleks dan metode Karmarkar. Penelitian ini dilakukan dengan studi literatur. Penyelesaian program linier dengan metode Karmarkar, mula-mula harus diubah dalam bentuk kanonik Karmarkar, kemudian diselesaikan dengan metode Karmarkar. Penyelesaian program linier dengan metode Karmarkar dilakukan secara manual dan dengan menggunakan program Matlab, kemudian hasil dari keduanya dilakukan analisis. Kesimpulannya adalah bahwa metode Karmarkar adalah suatu metode titik interior yang menembus dari daerah fisibel untuk mencapai suatu solusi optimum sedangkan metode simpleks bergerak dari titik ekstrim menuju ke penyelesain optimum. Titik interior dilambangkan dengan banyaknya variabel. Menyelesaikan masalah dengan metode Karmarkar yaitu dengan mengubah bentuk dasar program linier ke bentuk kanonik Karmarkar, dilanjutkan dengan perhitungan iterasi hingga nilai  minimum (kanonik Karmarkar kurang dari 0,05. Metode Karmarkar membutuhkan perhitungan yang relatif lebih besar untuk persoalan program linier yang berukuran kecil dan lebih cepat diselesaikan dengan metode simpleks, sedangkan untuk kendala yang lebih besar metode Karmarkar lebih efisien dibandingkan metode simpleks. This research purpose is to determine the basic mathematical Karmarkarmethods, to know the solving linear programs with Karmarkar method, and to analyze the problem solving linear program with the simplex method and Karmarkar method. This research was literature study. The completion of linear programs with Karmarkar method was done manually by using Matlab program, then the results of both was analyzed. The conclusion is the Karmarkar method is a method that penetrates the interior point of the feasible region to achieve an optimum solution while the simplex method

  4. Analisis Hubungan Ekspor, Impor, PDB dan Utang Luar Negeri Indonesia Periode 1970-2013

    Directory of Open Access Journals (Sweden)

    Dison M.H. Batubara

    2015-11-01

    Full Text Available Tujuan penelitian ini adalah untuk mengetahui ada tidaknya hubungan kausalitas serta kointegrasi di antara ekspor, impor, PDB dan utang luar negeri Indonesia dengan memakai data sekunder time series tahun 1970-2013. Penelitian ini menerapkan metode Vector Autoregression (VAR yang meliputi Granger-Causality test dan Johansen Co-Integration test, yang dilanjutkan dengan estimasi Vector Error Correction Model (VECM dan forecasting melalui analisis Impulse Response Function (IRF dan Forecast Error Variance Decomposition (FEVD. Hasil uji Granger-Causality menunjukkan diantara keempat variabel tidak terdapat kausalitas, namun terdapat lima hubungan satu arah (unidirectional, yang meliputi ekspor ke impor, ekspor ke utang luar negeri, PDB ke impor, impor ke utang luar negeri dan PDB ke utang luar negeri. Johansen Co-Integration test menunjukkan bahwa keempat variabel terkointegrasi. Analisis IRF dan FEVD menunjukkan bahwa variabel yang paling berpengaruh terhadap ekspor, impor dan PDB adalah ekspor, sedangkan variabel yang paling berpengaruh terhadap utang luar negeri adalah impor

  5. Analisi basata sugli sforzi locali della resistenza a fatica di giunzioni incollate di materiali compositi

    Directory of Open Access Journals (Sweden)

    F. Moroni

    2009-07-01

    Full Text Available Il lavoro prende spunto dai risultati di un’analisi sperimentale del comportamento a fatica di giunzioni incollate di materiali compositi laminati di elevato spessore formati da strati di unidirezionale e di tessuto di fibra di carbonio. I giunti sono stati realizzati in modo tale da saggiare l’influenza della lunghezza di sovrapposizione (da 25,4 mm a 110,8 mm, della forma del giunto (con e senza rastremazione, e della composizione degli aderendi (sostituzione di uno degli aderendi in composito con uno in acciaio. Mediante analisi 2D elastiche con il metodo degli elementi finiti sono state ricavate le distribuzioni degli sforzi all’interno dello strato di adesivo, al fine di individuare un parametro utile alla descrizione del comportamento a fatica in termini di sforzi locali - numero di cicli a rottura. Il ruolo della fase di propagazione viene discusso alla luce di osservazioni dell’avanzamento della frattura, condotta su alcuni dei giunti testati.

  6. Analisis perilaku dinamik pada sel T CD4+ dan sel T CD8+ terhadap infeksi mikobakterium

    Directory of Open Access Journals (Sweden)

    Alfi Nur Rochmatin

    2014-05-01

    Full Text Available Model matematika pada infeksi mikobakterium tuberkulosis yang berbentuk sistem persamaan diferensial nonlinear orde satu. Penelitian ini telah mengkonstruksi model matematika pada interaksi makrofag, sel T CD4+ dan sel T CD8+ dengan pengaruh usia. Solusi numerik pada model matematika ini dengan menggunakan ODE 45 berbantuan matlab. Analisis kestabilan diamati melalui titik tetap dengan mencari matriks jacobian dan nilai eigen dari titik tetap tersebut, maka dapat diperoleh bahwa semua titik tetap tersebut tidak stabil. Berdasarkan analisis perilaku dinamik pada sel T CD4+ dan sel T CD8+pada usia muda dan usia tua maka akan diperoleh bahwa sel T CD4+dan sel T CD8+ lebih banyak mempengaruhi populasi bakteri mikobakterium tuberkulosis dari pada saat usia muda

  7. Meta Analisis sa Pagsusuri ng Maiikling Kwento sa mga Tesis at Disertasyon

    OpenAIRE

    Regina I. Cuizon

    2014-01-01

    Anumang mga pagbabago na makikita sa mundo ay bunga ng pananaliksik. Ang meta-analisis ay isa sa mga bunga ng pananaliksik na ginagamit sa kasalukuyan bilang teknik upang malaman iba pang mga nagsulputang informasyon. Ito’y pag-aaral sa mga pag-aaral. Isang kritikal at sistematikong pagsusuri sa istruktura ng mga pag-aaral. Maging gabay sa mga gradwadong paaralan sa pagpili ng paksang pagaaralan. Pangunahing layunin na matiyak ang mga pamamaraan sa pagsusuri ng maiikling kwento ng...

  8. ANALISIS HEDONIC TERHADAP HARGA JUAL MOBIL TOYOTA KIJANG BEKAS DI SURABAYA

    Directory of Open Access Journals (Sweden)

    Guntur Martono

    2000-01-01

    Full Text Available The factors influencing the price of used Toyota Kijang vehicles in Surabaya were evaluated using a hedonic model. Regression analysis was used to estimate the size of the influence of the hedonic characterics which determine used Kijang prices. Based on the results, it can be concluded that the year the vehicle was made, the condition of the engine, the condition of the body, the condition of tires, the presence of air conditioner, and the GX and SX types all have a positive and significant relationship with the selling price. The value of R2 is 0,951 and the value of the F test is significant. The results of forecasting analysis show that the above analysis is correct, since the difference between the forecast price and the actual price varies between 1.2% - 4.5%. Abstract in Bahasa Indonesia : Faktor-faktor yang mempengaruhi harga mobil niaga Toyota Kijang bekas di Surabaya diteliti dengan menggunakan model hedonic. Analisis regresi digunakan untuk menghitung besarnya pengaruh karakteristik hedonic yang menentukan harga mobil Kijang bekas. Dari hasil analisis, dapat disimpulkan bahwa tahun pembuatan, kondisi mesin, kondisi bodi, kondisi ban, air conditioner, tipe GX dan tipe SX berhubungan positif dan signifikan terhadap harga jual. Nilai R2 menjadi sebesar 0,951 dan nilai uji F signifikan. Hasil analisis peramalan membuktikan bahwa analisis di atas tepat karena perselisihan antara harga peramalan dan harga sebenarnya berkisar antara 1,2% - 4,5%. Kata kunci: model hedonic, regresi berganda, variabel dummy, harga mobil bekas, Toyota Kijang

  9. Analisis Faktor-Faktor Yang Mempengaruhi Price Earning Ratio Pada Perusahaan LQ45 Di Indonesia

    OpenAIRE

    Hutagalung, Pretty N

    2010-01-01

    Pretty. N. Hutagalung (2010). Analisis Faktor-Faktor yang Mempengaruhi Price Earning Ratio pada Perusahaan LQ45 di Indonesia. Pembimbing, Dr. Isfenti Sadalia, SE, ME. Ketua Departemen, Prof. Dr. Ritha F. Dalimunthe, SE, M.Si. Penguji, Drs. Syahyunan, M.Si dan Drs. Nakman Harahap, Msi. (Penguji I dan Penguji II). Penelitian ini bertujuan untuk menguji dan menganalisis pengaruh Dividend Payout Ratio, Earning Growth, Debt to Equity Ratio dan Return on Equity terhadap Price Earning Ratio pada...

  10. Analisis Perkembangan Usaha Home Industry Makanan dan Minuman di Kota Binjai

    OpenAIRE

    Muharoma, Fadhlun Wirizkho

    2016-01-01

    This research which titled "Analisis Perkembangan Usaha Home Industry Makanan dan Minuman di Kota Binjai " aims to determine the internal factors that consist of strengths and weaknesses, as well as external factors which consist of the opportunities and threats of home industry business development strategies in Binjai. The data used are primary data and secondary data. Primary data were obtained from questionnaires by speakers, namely home business operators in the food and beverage industr...

  11. REPRESENTASI EKSISTENSI DIRI PADA PROFILE PICTURE DALAM SITUS PERTEMANAN FACEBOOK (Sebuah Analisis Semiotika)

    OpenAIRE

    LA???LANG, RAKHMAWATY

    2012-01-01

    RAKHMAWATY LA???LANG, E31107072. Representasi Eksistensi Diri pada Profile Picture dalam Situs Pertemanan Facebook (Analisis Semiotika). (dibimbing M. Farid dan Alem Febri Sonni). Tujuan penelitian ini adalah: 1. Untuk mengetahui bagaimana eksistensi seseorang dihadirkan dalam Profile Picture pada situs pertemanan melalui penelitian terhadap semua unsur tanda yaitu ikon, indeks, dan simbol. 2. Untuk mengetahui representasi eksistensi diri dalam Profile Picture pada Facebook. Penelitian me...

  12. Analisis Semiotika Makna Pesan pada Iklan Axis Versi “Iritologi – Menatap Masa Depan”

    OpenAIRE

    Puri Sulistiyawati

    2016-01-01

    Abstrak Tujuan dari penelitian ini adalah untuk mengetahui dan menganalisis makna pesan dalam iklan Axis versi “Iritologi – Menatap masa depan” di televisi dengan menggunakan pendekatan semiotika Roland Barthes. Dari analisis yang telah dilakukan berdasarkan makna denotasi, konotasi dan mitos dalam iklan Axis versi “iritologi – Menatap masa depan”, dapat diketahui bahwa iklan Axis memiliki konsep yang sangat sederhana, dengan mengusung tema tentang kebiasaan remaja dalam memanfaatkan media s...

  13. Productividad del trabajo en la industria maquiladora del norte de Mexico: Un analisis de convergencia

    OpenAIRE

    Jorge Eduardo Mendoza Cota

    2004-01-01

    En este estudio se analiza la productividad del trabajo en la industria maquiladora de exportacion por divisiones y por estados, comparando la region de la frontera norte de Mexico con el resto de los estados. Se utiliza un analisis econometrico de panel, partiendo del enfoque de convergencia condicional. Se aprecia una heterogeneidad de la productividad del trabajo, lo que sugiere diferencias regionales en los procesos tecnologicos y en la dotacion del capital. Los estados del centro mostrar...

  14. L’analisi dell’attività onirica secondo l’approccio cognitivo - neuropsicologico

    Directory of Open Access Journals (Sweden)

    Nicola Allegri

    2012-09-01

    Full Text Available Un’analisi fenomenologica del sogno apre il quadro delle indagini scientifiche sull’attività onirica e consente a differenti discipline di integrare le rispettive conoscenze empiriche sull’argomento. In questo lavoro, a partire da una visione fenomenologica ed ermeneutica dell’essere umano (essere-nel-mondo, cercheremo di delineare alcuni aspetti della struttura onirica, facendo dialogare tra loro alcune evidenze empiriche sull’argomento che provengono sia dalle scienze umane, sia dalle neuroscienze.

  15. ANALISIS FISIS MEMBRAN BIOFILTER ROKOK DENGAN VARIASI DAUN, BIJI DAN KULIT DELIMA

    OpenAIRE

    Umaiyatus Syarifah; Ririn Mega S; Muthmainnah; Agus Mulyono

    2015-01-01

    Analisis fisis membrane biofilter rokok telah dilakukan untuk mengetahui kerapatan dan porositas. Membrane biofilter terbuat dari variasi daun delima, biji delima dan kulit delima. Variasi massa yang digunakan pada masing-masing bahan delima adalah 0.7 gram, 0.8 gram, 0.9 gram dan 1 gram. Matriks yang digunakan pada pembuatan biofilter berbahan delima adalah polyethilene glikol (PEG). Data pengujian kerapatan membran pada biofilter berbahan daun delima terbesar adalah 1.532 g/cm3 dengan kompo...

  16. Analisis Customer Relationship Management Terhadap Kepuasan Nasabah Pada Customer Care Bank CIMB Niaga Makassar

    OpenAIRE

    Ihsan Raya, Ahmad

    2013-01-01

    2013 Penelitian ini bertujuan untuk mengetahui pengaruh Customer Relationship management Continuity marketing, one to one marketing, dan partnering atau co-marketing terhadap kepuasan nasabah pada Customer Care Bank CIMB Niaga Makassar dan untuk mengetahui variabel dari Customer Relationship yang paling dominan berpengaruh terhadap kepuasan nasabah pada Customer Care Bank CIMB Niaga Makassar. Untuk mencapai tujuan tersebut maka digunakan metode analisis deskriptif, uji kelayakan instrument...

  17. ANALISIS KINERJA KEUANGAN PT BANK SULSELBAR DENGAN MENGGUNAKAN ECONOMIC VALUE ADDED (EVA)

    OpenAIRE

    Asmiati, -

    2012-01-01

    ASNIATI, A21108323. Analisis Kinerja Keuangan PT Bank Sul-Selbar Dengan Menggunakan Economic Value Added (EVA). (Dibimbing Oleh Prof.Dr.H Syamsu Alam SE, M.si dan Drs Mukhtar M.si) PT Bank Sul-selbar merupakan perusahaan perbankan yang Menjadi Bank terbaik di Kawasan Indonesia Timur dengan dukungan manajemen dan sumber daya manusia yang profesional serta memberikan nilai tambah kepada Pemda dan masyarakat. Penelitian ini ber...

  18. Analisis Pengaruh Ekspor Sektor Industri Dan Penanaman Modal Asing Sektor Industri Terhadap Pertumbuhan Ekonomi Indonesia

    OpenAIRE

    Fahmi Hasbullah

    2009-01-01

    Penelitian ini berjudul “Analisis Pengaruh Ekspor Sektor Industri dan Penanaman Modal Asing Sektor Industri terhadap Pertumbuhan Ekonomi Indonesia”. Dalam kasus ini, Pertumbuhan Ekonomi Indonesia adalah variabel terikat. Perdagangan Internasional yang terdiri dari Ekspor dan Impor adalah variabel bebas. Tujuan penelitian ini adalah untuk menjelaskan pengaruh variabel bebas terhadap variabel terikat. Penelitian ini menggunakan data sekunder atau data periode waktu sejak 1987 sampai 2006. Da...

  19. Analisis Kata Penghubung He (和), Gen ( 跟) Dan Yu (与) Dalam Kalimat Bahasa Mandarin

    OpenAIRE

    Tarigan, Cicilia Aprilina Kartika

    2013-01-01

    The tittle of this paper is “Analisis Kata Penghubung He, Gen, dan Yu dalam Kalimat Bahasa Mandarin. Researcher analyzes the use of Chinese cunjuctions, he, gen, yu in Chinese sentences. Actually he, gen and yu have the same meaning but differently in their own characteristics. The concepts of the thesis are The methodology on the thesis is descriptive method. The author analyze about characteristic and similarities and differences of cunjuctions, hen, gen, and yu. By this, author is try to ...

  20. Analisis Penempatan Terhadap Kinerja Karyawan pada PT. Perkebunan Nusantara III Medan

    OpenAIRE

    Yuni Ramadhani Nasution

    2008-01-01

    ABSTRAK Yuni Ramadhani Nst (2006), Analisis Penempatan Terhadap Kinerja Karyawan pada PT. Perkebunan Nusantara III Medan, dibawah bimbingan Prof. Dr. Ritha F. Dalimunthe, SE, MSi selaku Ketua Departemen, Drs. Chairuddin Nst selaku Dosen Pembimbing, Dra, Ramona RI Hasibuan, MRP selaku Dosen Penguji I dan Dra, Nurbayati Siregar, MSi selaku Dosen Penguji II. Penelitian ini bertujuan mengetahui adanya pengaruh komunikasi terhadap kinerja karyawan pada PT. Perkebunan Nusantara III Medan . Pe...