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Sample records for nordostschweizerische kraftwerk-2 reaktor

  1. ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI

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    Azizul Khakim

    2015-10-01

    Full Text Available ABSTRAK ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI. Bulk shielding merupakan fasilitas yang terintegrasi dengan reaktor Kartini yang berfungsi sebagai penyimpanan sementara bahan bakar bekas. Fasilitas ini merupakan fasilitas yang termasuk dalam struktur, sistem dan komponen (SSK yang penting bagi keselamatan. Salah satu fungsi keselamatan dari sistem penanganan dan penyimpanan bahan bakar adalah mencegah kecelakaan kekritisan yang tak terkendali dan membatasi naiknya temperatur bahan bakar. Analisis keselamatan paling kurang harus mencakup analisis keselamatan dari sisi neutronik dan termo hidrolik Bulk shielding. Analisis termo hidrolik ditujukan untuk memastikan perpindahan panas dan proses pendinginan bahan bakar bekas berjalan baik dan tidak terjadi akumulasi panas yang mengancam integritas bahan bakar. Code tervalidasi PARET/ANL digunakan untuk analisis pendinginan dengan mode konveksi alam. Hasil perhitungan menunjukkan bahwa mode pendinginan konvekasi alam cukup memadai dalam mendinginkan panas sisa tanpa mengakibatkan kenaikan temperatur bahan bakar yang signifikan. Kata kunci: Bulk shielding, bahan bakar bekas, konveksi alam, PARET.   ABSTRACT THERMAL HYDRAULIC SAFETY ANALYSIS OF BULK SHIELDING KARTINI REACTOR. Bulk shielding is an integrated facility to Kartini reactor which is used for temporary spent fuels storage. The facility is one of the structures, systems and components (SSCs important to safety. Among the safety functions of fuel handling and storage are to prevent any uncontrolable criticality accidents and to limit the fuel temperature increase. Safety analyses should, at least, cover neutronic and thermal hydraulic calculations of the bulk shielding. Thermal hydraulic analyses were intended to ensure that heat removal and the process of the spent fuels cooling takes place adequately and no heat accumulation that challenges the fuel integrity. Validated code, PARET/ANL was used for analysing the

  2. REAKTOR INNOVATIVE MOLTEN SALT (IMSR DENGAN SISTEM KESELAMATAN PASIF MENYELURUH

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    Andang Widiharto

    2015-04-01

    Full Text Available Pengembangan Teknologi Reaktor Nuklir pada masa mendatang mengarah pada peningkatan aspek keselamatan, peningkatan pendayagunaan bahan bakar, reduksi limbah radioaktif, ketahanan terhadap proliferasi bahan-bakar nuklir dan peningkatan aspek ekonomi. reaktor Innovative Molten Salt (IMSR adalah reaktor nuklir yang menggunakan bahan bakar cair berupa garam lebur fluoride (7LiF-ThF4-UF4-MaFx. Reaktor IMSR didesain sebagai reaktor pembiak termal, yaitu membiakkan U-233 dari Th-232. Hal ini untuk menjawab permasalahan sustainabilitas ketersedian sumber daya bahan bakar nuklir dan reduksi limbah radioaktif. Dalam aspek keselamatan, desain reaktor IMSR memiliki sifat inherent safe, yaitu koefisien umpan balik daya yang negatif serta memiliki fitur-fitur keselamatan pasif. Fitur-fitur keselamatan pasif terdiri dari sistem shutdown pasif, sistem pendinginan pasif pasca shutdown serta sistem pendinginan pasif untuk produk fisi. Kecelakaan yang berpotensi terjadi pada IMSR, yaitu kecelakaan kehilangan aliran bahan bakar, kecelakaan kehilangan aliran pendingin, kecelakaan kehilangan kemampuan pengambilan kalor serta kecelakaan kerusakan integritas sistem reaktor, dapat ditangani sepenuhnya secara pasif hingga mencapai kondisi shutdown selamat. Kata kunci: keselamatan pasif, inherent safe, IMSR   The next Nuclear Reactor Technology developments are directed to the increasing of the aspects of safety, fuel utility, radioactive waste reduction, proliferation retention and economy. Innovative Molten Salt Reactor (IMSR is a nuclear reactor design that uses fluoride molten salt (7LiF-ThF4-UF4-MaFx. IMSR is designed as a thermal breeder reactor, i.e. to produce U-233 from Th-232. This is the answer of natural nuclear fuel sustainability and radioactive waste problems. In term of safety aspect, IMSR design has inherent safe characteristics, i.e. negative power feedback coefficient, and passive safety features. The passive safety features are passive shutdown

  3. ANALISIS PENGENDALIAN DAYA REAKTOR PCMSR DENGAN LAJU ALIR PENDINGIN

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    Iqbal Syafin Noha

    2015-03-01

    Full Text Available Passive Compact Molten Salt Reactor (PCMSR merupakan pengembangan dari Molten Salt Reactor (MSR yang memiliki karakter berbeda dengan lima reaktor generasi IV lainnya, yaitu menggunakan bahan bakar leburan garam. Pada reaktor MSR, garam lebur tidak digunakan sebagai pendingin tetapi digunakan sebagai medium pembawa bahan bakar. Dengan fase bahan bakar yang berupa garam lebur LiF-BeF2-ThF4-UF4, maka dapat dilakukan pengendalian daya dengan mengatur laju aliran bahan bakar dan pendingin. Tujuan penelitian ini adalah untuk mengetahui pengaruh perubahan laju alir pendingin terhadap daya reaktor PCMSR. Analisis dilakukan dengan empat jenis masukan untuk perubahan laju alir pendingin, yaitu masukan step, ramp, eksponensial, dan sinusoidal. Untuk masukan step, laju alir pendingin dibuat berubah secara mendadak. Selanjutnya untuk masukan ramp dan eksponensal, perubahan laju alir masing-masing dibuat perlahan secara linear dan mengikuti fungsi eksponensial. Kemudian untuk masukan sinusoidal, laju alir berubah naik turun secara periodik dengan memvariasikan frekuensi dari perubahan laju alir tersebut. Hasil penelitian menunjukkan bahwa penurunan laju alir pendingin sebesar 50% dari laju pendingin sebelumnya, menyebabkan daya pada reaktor PCMSR turun sebesar 63% dari daya sebelumnya. Jika terjadi fluktuasi laju aliran pendingin, maka semakin cepat perubahan tersebut, maka respon daya yang diberikan semakin kecil. Pada frekuensi yang sangat cepat, daya reaktor menjadi konstan dan cenderung tidak memiliki respon terhadap laju aliran. Hal ini merupakan salah satu aspek keselamatan reaktor, karena reaktor tidak merespon perubahan yang terlalu cepat. Kemampuan reaktor mengatur daya menyesuaikan laju aliran pendingin merupakan aspek keselamatan lainnya. Kata kunci : PCMSR, pengendalian daya, laju alir pendingin, uji respon   Passive Compact Molten Salt Reactor (PCMSR is the development of Molten Salt Reactor (MSR which has different character from other five

  4. Fire modeling of the Heiss Dampf Reaktor containment

    International Nuclear Information System (INIS)

    Nicolette, V.F.; Yang, K.T.

    1995-09-01

    This report summarizes Sandia National Laboratories' participation in the fire modeling activities for the German Heiss Dampf Reaktor (HDR) containment building, under the sponsorship of the United States Nuclear Regulatory Commission. The purpose of this report is twofold: (1) to summarize Sandia's participation in the HDR fire modeling efforts and (2) to summarize the results of the international fire modeling community involved in modeling the HDR fire tests. Additional comments, on the state of fire modeling and trends in the international fire modeling community are also included. It is noted that, although the trend internationally in fire modeling is toward the development of the more complex fire field models, each type of fire model has something to contribute to the understanding of fires in nuclear power plants

  5. DESAIN KONSEPTUAL TERAS REAKTOR RISET INOVATIF BERBAHAN BAKAR URANIUM MOLIBDENUM DARI ASPEK NEUTRONIK

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    Tukiran Surbakti

    2015-03-01

    Full Text Available Manfaat yang luas dari penggunaan reaktor riset membuat banyak negara membangun reaktor riset baru. Kecenderungan saat ini adalah reaktor tipe reaktor serbaguna (MPR dengan teras yang kompak untuk mendapatkan fluks neutron yang tinggi dengan daya yang relatif sedang atau rendah. Reaktor riset yang ada di Indonesia yang paling muda usianya sudah berumur 25 tahun. Oleh karena itu diperlukan desain reaktor riset baru sebagai alternatif, disebut reaktor riset inovatif (RRI, kelak pengganti reaktor riset yang sudah ada. Tujuan dari riset ini mendapatkan konfigurasi teras setimbang reaktor riset yang optimal dengan kriteria memiliki fluks neutron termal minimum sebesar 2,5x1014 n/cm2 s pada daya 20 MW (minimum, memiliki panjang operasi satu siklus lebih dari 40 hari dan penggunaan bahan bakar yang paling efisien. Desain neutronik dilakukan untuk bahan bakar baru U-9Mo-Al dengan kerapatan bervariasi dan jenis reflektor yang bervariasi. Desain dilakukan dengan paket program WIMSD-5B dan BATAN-FUEL. Hasil desain konseptual menyajikan 4 konfigurasi teras yaitu 5×5, 5×7, 6×5 dan 6×6. Hasil optimasi menunjukkan bahwa teras setimbang reaktor RRI dengan konfigurasi 5×5, tingkat muat 235U sebesar 450 g, reflektor berilium, fluks neutron termal maksimum di daerah reflektor sebesar 3,33×1014 neutron cm-2s-1 dan panjang siklus 57 hari merupakan desain teras reaktor riset inovatif yang paling optimal. Kata kunci: desain konseptual, bahan bakar uranium-molibdenum,berilium, D2O, WIMS, BATAN-FUEL   The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR and then as an alternative to replace the old

  6. RA research nuclear reactor - Annual report 1985; Istrazivacki nuklearni reaktor RA - Izvestaj za 1985. godinu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-12-01

    Research reactor RA Annual report for year 1985 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1985 godinu sastoji se od dva dela: prvi deo obuhvata pogon i odzavanje reaktora RA, a drugi poslove zastite od zracenja na reaktoru RA.

  7. RA Research reactor, Annual report 1986; Istrazivacki nuklearni reaktor RA - Izvestaj za 1986. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1986-12-01

    Research reactor RA Annual report for year 1985 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1986 godinu sastoji se od dva dela: prvi deo obuhvata pogon i odzavanje reaktora RA, a drugi poslove zastite od zracenja na reaktoru RA.

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

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    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.

  9. ANALISIS PENGARUH DENSITAS BAHAN BAKAR SILISIDA TERHADAP PARAMETER KINETIK TERAS REAKTOR RSG-GAS

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    Tukiran s

    2016-11-01

    Full Text Available Saat ini RSG-GAS menggunakan elemen bakar silisida 2,96 g U/cc. Untuk meningkatkan waktu operasi reaktor maka akan direncanakan untuk mengganti elemen bakar silisida dengan kerapatan yang lebih tinggi. Keuntungan reaktor dengan bahan bakar kerapatan tinggi adalah dapat lebih efektif dan efisien. Maka perlu dilakukan perhitungan parameter kinetik teras silisida kerapatan tinggi mengingat pengaruhnya sangat penting untuk keselamatan operasi reaktor. Parameter kinetik yang dihitung yaitu fraksi neutron kasip efektif, konstanta peluruhan neutron kasip, umur neutron serempak yang merupakan faktor utama dalam kontrol dan keselamatan. Bahan bakar silisida tipe pelat dengan densitas 2,96 - 4,8 gU/cm3 digunakan pada teras RSG-GAS untuk menganalisis perhitungan parameter kinetik. Perhitungan sel dilakukan dengan paket program WIMSD-5B dan paket program Batan-2DIFF digunakan untuk perhitungan teras. Hasil perhitungan menunjukkan bahwa harga fraksi neutron kasip turun dengan naiknya densitas bahan bakar. Turunnya nilai parameter kinetik ini tidak mengganggu pergantian bahan bakar ke densitas yang lebih tinggi. Turunnya nilai parameter kinetik rata-rata dari densitas 2,96 gU/cm3 ke 3,55 gU/cm3 adalah 1,3 % sedangkan dari densitas 2,96 gU/cm3 ke 4,8 gU/cm3 adalah 2,2 % . Sehingga jika dilakukan pergantian bahan bakar maka ditinjau dari segi neutronik dan parameter kinetiknya tidak akan mengalami perubahan dalam pola operasi reaktor atau manajemen bahan bakar dan tidak akan berpengaruh terhadap keselamatan operasi reaktor.

  10. ANALISIS POLA MANAJEMEN BAHAN BAKAR DESAIN TERAS REAKTOR RISET TIPE MTR

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    Lily Suparlina

    2015-03-01

    Full Text Available Parameter neutronik dibutuhkan dalam mendesain teras reaktor riset. Reaktor riset jenis MTR (Material Testing Reactor sangat diminati karena dapat digunakan baik untuk riset dan juga produksi radio isotop. Reaktor riset yang ada saat ini sudah tua sehingga dibutuhkan desain reaktor yang mempunyai teras kompak. Desain teras reaktor riset yang sudah ada saat ini belum cukup memadai untuk memenuhi persyaratan di dalam UCD yang telah ditetapkan yaitu fluks neutron termal di teras 1x1015 n/cm2s, oleh karena itu perlu dibuat desain teras reaktor baru sebagai alternatif yang kompak dan dapat menghasilkan fluks neutron tinggi. Telah dilakukan perhitungan dan analisis terhadap manajemen bahan bakar desain teras kompak dengan konfigurasi teras 5x5, berbahan bakar U9Mo-Al dan tinggi teras aktif 70 cm. Tujuan dari riset ini untuk memperoleh fluks neutron di teras memenuhi kebutuhan seperti yang telah ditetapkan di UCD dengan panjang siklus operasi minimum 20 hari pada daya 50 MW. Perhitungan dilakukan dengan menggunakan paket program komputer WIMSD-5B untuk menggenerasi tampang lintang makroskopik bahan bakar dan Batan-FUEL untuk memperoleh nilai parameter neutronik serta Batan-3DIFF untuk perhitungan nilai reaktivitas batang kendali. Perhitungan parameter neutronik teras reaktor riset ini dilakukan untuk bahan bakar U-9Mo-Al dengan tingkat muat bervariasi dan 2 macam pola pergantian bahan bakar yaitu teras segar dan teras setimbang. Hasil analisis menunjukkan bahwa pada teras segar, tingkat muat 235U sebesar 360 gram, 390 gram dan 450 gram memenuhi kriteria keselamatan dan kriteria penerimaan di UCD dengan nilai fluks neutron termal di teras lebih dari 1x1015 n/cm2s dan panjang siklus >20 hari, sedangkan pada teras setimbang panjang siklus dapat terpenuhi hanya untuk tingkat muat 450 gram. Kata kunci: desain teras reaktor, bahan bakar UMo, pola bahan bakar, WIMS, BATAN-FUEL   Research reactor core design needs neutronics parameter calculation use computer

  11. ANALISIS SEBARAN RADIONUKLIDA PADA KONDISI NORMAL UNTUK REAKTOR AEC 1000 MW

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    Sri Kuntjoro

    2015-03-01

    Full Text Available Telah dilakukan analisis sebaran radionuklida pada reaktor daya Atomic Energy Agency (AEC 3568 MWTh, setara dengan 1000 Mwe untuk kondisi operasi normal. Analisis dilakukan untuk dua reaktor yang terpisah sejauh 500 m dan sudut 90o satu dengan yang lain. Langkah awal dalam melakukan analisis adalah menentukan suku sumber reaktor menggunakan program komputer ORIGEN2 dan EMERALD NORMAL. ORIGEN2 digunakan untuk menentukan inventori radionuklida yang terdapat di reaktor. Selanjutnya dengan dengan menggunakan program EMERALD NORMAL dihitung suku sumber yang sampai ke cerobong reaktor. Untuk menganalisis dosis yang diterima penduduk dilakukan dengan menggunakan program PC-CREAM. Perhitungan dilakukan untuk satu dan dua PLTN di calon tapak PLTN. Hasil yang diperoleh adalah sebaran radionuklida terbesar untuk satu PLTN pada jarak 1 km dan kearah zona 9 (191,25o dan untuk dua PLTN pada jarak 1 km dan kearah zona 10 (213,75o. Radionuklida yang sampai ke penduduk melalui dua alur yaitu alur makanan dan hirupan. Untuk alur makanan berasal dari radionuklida I-131, dan terbesar melalui alur produk susu sebesar 53,40 % untuk satu maupun dua PLTN . Untuk alur hirupan ranionuklida pemberi kontribusi paparan terbesar berasal dari Kr-85m sebesar 53,80 %. Dosis total terbesar yang diterima penduduk terdapat pada jarak 1 Km untuk bayi yaitu sebesar 4,10 μSi dan 11,26 μSi untuk satu dan dua PLTN. Hasil ini sangat kecil dibandingkan dengan batas dosis yang diijinkan oleh badan pengawas (BAPETEN untuk penduduk yaitu sebesar 1 mSi. Kata Kunci : Reaktor daya, komputer code, radionuklida, alur makanan, hirupan   Analysis for radionuclide dispersion for the Atomic Energy Agency (AEC 3568 MWth Power Reactor, equal to the 1000 MWe at normal condition has been done. Analysis was done for two piles that is separated by 500 m distance and angle of 90o one to other. Initial pace in doing the analysis is to determine reactors source term using ORIGEN2 and EMERALD NORMAL

  12. ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA

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    Andi Sofrany Ekariansyah

    2015-06-01

    Full Text Available ABSTRAK ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA. Kecelakaan yang diakibatkan oleh kehilangan pendingin (loss of coolant accident / LOCA dari sistem reaktor merupakan kejadian dasar desain yang tetap diantisipasi dalam desain reaktor daya yang mengadopsi teknologi Generasi II hingga IV. LOCA ukuran kecil (small break LOCA memiliki dampak yang lebih signifikan terhadap keselamatan dibandingkan LOCA ukuran besar (large break LOCA seperti terlihat pada kejadian Three-Mile Island (TMI. Fokus makalah adalah pada analisis small break LOCA pada reaktor daya maju Generasi III+ yaitu AP1000 dengan mensimulasikan tiga kejadian pemicu yaitu membukanya katup Automatic Depressurization System (ADS secara tak disengaja, putusnya salah satu pipa Direct Vessel Injection (DVI secara double-ended, dan putusnya pipa lengan dingin dengan diameter bocoran 10 inci. Metode yang digunakan adalah simulasi kejadian pada model AP1000 yang dikembangkan secara mandiri menggunakan program perhitungan RELAP5/SCDAP/Mod3.4. Dampak yang ingin dilihat adalah kondisi teras selama terjadinya small break LOCA yang terdiri dari pembentukan mixture level dan transien temperatur kelongsong bahan bakar. Hasil simulasi menunjukkan bahwa mixture level untuk semua kejadian small break LOCA berada di atas tinggi teras aktif yang menunjukkan tidak terjadinya core uncovery. Adanya mixture level berpengaruh pada transien temperatur kelongsong yang menurun dan menunjukkan pendinginan bahan bakar yang efektif. Hasil di atas juga identik dengan hasil perhitungan program lain yaitu NOTRUMP. Keefektifan pendinginan teras juga disebabkan oleh berfungsinya injeksi pendingin melalui fitur keselamatan pasif yang menjadi ciri reaktor daya AP1000. Secara keseluruhan, hasil analisis menunjukkan model AP1000 yang telah dikembangkan dengan RELAP5 dapat digunakan untuk keperluan analisis kecelakaan dasar desain pada reaktor daya maju AP1000. Kata kunci: analisis

  13. ANALISIS DESAIN PROSES SISTEM PENDINGIN PADA REAKTOR RISET INOVATIF 50 MW

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    Sukmanto Dibyo

    2015-03-01

    Full Text Available Reaktor Riset Inovatif (RRI merupakan jenis MTR (Material Testing Reactor yang dipersiapkan ke depan sebagai desain reaktor baru. Daya RRI telah ditetapkan dari perhitungan neutronik dan termohidrolika teras yaitu 50 MW termal. Reaktor bertekanan 8 kgf/cm2 dan laju aliran massa pendingin primer 900 kg/s. Tantangan yang penting dalam menindak lanjuti desain reaktor ini adalah analisis desain pada sistem pendingin. Makalah ini bertujuan untuk menganalisis desain proses sistem pendingin utama reaktor RRI daya 50 MW (RRI-50 dengan menggunakan program Chemcad 6.1.4. Dalam analisis ini dilakukan perhitungan neraca massa dan energi (mass/energy balances pada sistem pendingin primer dan sekunder sebagai pendingin utama. Masing-masing sistem pendingin tersebut terdiri dari 2 jalur beroperasi secara paralel dan 1 jalur redundansi. Disamping itu untuk desain termal unit komponen telah dianalisis dengan program RELAP5, frenchcreek dan Metoda Analitik. Hasil analisis yang diperoleh adalah desain diagram sistem pendingin yang mencakup data parameter entalpi, temperatur, tekanan dan laju aliran massa pendingin untuk masing-masing jalur. Adapun hasil desain unit komponen utama pada RRI-50 adalah tangki tunda dengan volume 51,5 m3, 2 unit pompa sentrifugal dan 1 unit pompa cadangan pada pendingin primer daya 141 kW/pompa dan pendingin sekunder daya 206 kW/pompa, 2 unit penukar panas tipe shell-tube dengan koefisien termal overall 1377 W/m2.oC dan 4 unit menara pendingin yang mampu melepaskan panas ke udara dengan desain temperatur approach 5,0 oC dan temperatur range 9,0 oC. Desain sistem pendingin reaktor RRI-50 ini telah menetapkan parameter operasi sistem pendingin yaitu temperatur, tekanan dan laju aliran massa pendingin dengan mempertimbangkan tuntutan aspek keselamatan teras reaktor sehingga desain temperatur maksimum pendingin masuk ke teras 44,5 oC. Kata kunci : RRI 50 MW, desain sistem pendingin, program Chemcad 6.1.4   Innovative Research Reactor RRI

  14. PEMODELAN TERAS UNTUK ANALISIS PERHITUNGAN KONSTANTA MULTIPLIKASI REAKTOR HTR-PROTEUS

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    Zuhair Zuhair

    2015-04-01

    Full Text Available PTRKN sebagai salah satu unit kerja di BATAN dengan tugas pokok dan fungsi yang berkaitan erat dengan teknologi reaktor dan keselamatan nuklir, menaruh perhatian khusus pada konsep reaktor pebble bed. Dalam makalah ini pemodelan reaktor pebble bed HTR-PROTEUS dilakukan dengan program transport Monte Carlo MCNP5. Partikel bahan bakar berlapis TRISO dimodelkan secara detail dan eksak dimana distribusi acak partikel ini dalam bola bahan bakar didekati menggunakan array teratur kisi SC dengan fraksi packing 5,76% tanpa zona eksklusif. Model teras pebble bed didekati dengan memanfaatkan kisi teratur dari bola yang disusun sebagai kisi BCC berdasarkan sel berulang yang digenerasi dari sejumlah sel satuan. Hasil perhitungan MCNP5 memperlihatkan kesesuaian yang sangat baik dengan eksperimen, walaupun teras HTR-PROTEUS diprediksi lebih reaktif daripada pengukuran, khususnya di teras 4.2 dan 4.3. Pustaka ENDF/B-VI menunjukkan konsistensi dengan estimasi keff paling akurat dibandingkan pustaka ENDF/B-V, terutama ENDF/B-VI (66c. Deviasi estimasi keff yang dihitung dengan eksperimen dikaitkan sebagai konsekuensi dari komposisi reflektor grafit yang dispesifikasikan. Komparasi yang dibuat memperlihatkan bahwa MCNP5 menghasilkan keff teras HTR-PROTEUS lebih presisi daripada hasil dari MCNP4B dan MCNPBALL. Hasil ini menyimpulkan bahwa, sukses metodologi pemodelan ini menjustifikasi aplikasi MCNP5 untuk analisis reaktor pebble bed lainnya. Kata kunci: pemodelan teras HTR-PROTEUS, konstanta multiplikasi, MCNP5   PTRKN as a working unit in BATAN whose main duties and functions are related to reactor technology and nuclear safety, consern attention to pebble bed reactor concept. In this paper modeling of HTR-PROTEUS pebble bed reactor was done using Monte Carlo transport code MCNP5. The TRISO coated fuel particle is modeled in detailed and exact manner where random distributions of these particles in fuel pebble is approximated by using regular array of SC lattice

  15. Disain Sistem Pemantauan Lingkungan Untuk Evaluasi Lepasan Radionuklida dari Subsistem pada Kecelakaan Reaktor Daya PWR

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    Sri Kuntjoro

    2013-03-01

    Full Text Available PLTN. (Pembangkit Listrik Tenaga Nuklir sebagai sumber energi baru dipilih sebagai alternatif, karena memiliki berbagai kelebihan yaitu ramah lingkungan, pasokan bahan bakar yang tidak bergantung musim, serta harganya yang dapat bersaing dengan pembangkit listrik yang lain. Namun demikian, adanya keraguan sebagian masyarakat tentang keselamatan radiasi PLTN, maka pemerintah harus bisa meyakinkan tentang operasi PLTN yang aman dan selamat. Penelitian tentang disain sistem pemantauan lingkungan untuk evaluasi lepasan radionuklida dari subsistem reaktor dan lingkungan akibat terjadinya kecelakaan pada reaktor daya telah dilakukan. Penelitian dilakukan dengan melakukan perhitungan sebaran radionuklida ke subsistem dan lingkungan serta membuat sistim monitoring radiasi di lingkungan. Sistem monitoring lingkungan terdiri dari system pencacah radiasi, sistem peringatan dini, sistem pengukuran meteorologi, sistem GPS dan system GIS. Sistem pencacah radiasi digunakan untuk mencatat data radiasi, sistem pengukuran meteorologi digunakan untuk mencatat data arah dan kecepatan angin, sedangkan sistem GPS digunakan untuk menentukan data posisi pengukuran. Data tersebut kemudian dikirimkan ke system akuisisi data untuk ditransmisikan ke pusat kendali. Pengumpulan dan pengiriman data dilakukan melalui SMS menggunakan perangkat modem yang ditempatkan di ruang kendali. Ruang kendali menerima data dari berbagai tempat pengukuran. Dalam hal ini ruang kendali memiliki fungsi sebagai SMS gateway. Sistem ini dapat memvisualisasi untuk lokasi pengukuran yang berbeda. Selanjutnya, data posisi dan data radiasi diintegrasikan dengan peta digital. Integrasi sistem tersebut kemudian divisualisasikan dalam personal komputer. Untuk posisi pengukuran terlihat langsung di peta dan untuk data radiasi ditampilkan di monitor dengan tanda lingkaran merah atau hijau yang digunakan sebagai pemonitor batas aman radiasi. Bila tanda lingkaran berwarna merah maka akan menyalakan alarm di

  16. Isotopic composition of precipitation at the station Ljubljana (Reaktor, Slovenia – period 2007–2010

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    Polona Vreča

    2014-12-01

    Full Text Available The stable isotopic composition of hydrogen and oxygen (δ2H and δ18O and the tritium activity (A were monitored in monthly collected precipitation at Ljubljana (Reaktor during the period 2007–2010. Monthly and yearly isotope variations are discussed and compared with those observed over the period 1981–2006 and with the basic meteorological parameters for Ljubljana (Bežigrad and Ljubljana (Hrastje stations for the period 2007−2010. The mean values for δ2H and δ18O, weighted by precipitation amount at Ljubljana (Reaktor, are –59.4 ‰ and –8.71 ‰. The reduced major axis local meteoric water line (LMWLRMA is δ2H = (8.19 ± 0.22×δ18O + (11.52 ± 1.97, while the precipitation weighted least square regression results in LMWLPWLSR-Re δ2H = (7.94 ± 0.21×δ18O + (9.76 ± 1.93. The lack of significant difference in the LMWL slopes indicates a relatively homogeneous distribution of monthly precipitation as well as the small number of low-amount monthly precipitation events with low deuterium excess. The deuterium excess weighted mean value is 10.3 ‰ which indicates the prevailing influence of the Atlantic air masses. The temperature coefficient of δ18O is 0.30 ‰/°C. Tritium activity in monthly precipitation shows typical seasonal variations, with a weighted mean tritium activity in this period of 8.5 TU. No decrease of mean annual activity is observed.

  17. PEMBUATAN SERBUK U-6Zr DENGAN PENGKAYAAN URANIUM 19,75 % UNTUK BAHAN BAKAR REAKTOR RISET

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    Masrukan Masrukan

    2016-03-01

    Full Text Available ABSTRAK PEMBUATAN SERBUK PADUAN U-6Zr DENGAN PENGKAYAAN URANIUM 19,75 % UNTUK BAHAN BAKAR REAKTOR RISET. Telah dilakukan pembuatan serbuk paduan U-6Zr dengan pengkayaan 19,75 % untuk bahan bakar reaktor riset. Pembuatan bahan bakar U-6Zr ini dalam rangka mencari bahan bakar baru yang mempunyai densitas tinggi untuk mengganti bahan bakar yang sudah ada U3Si2-Al. Tujuan dari percobaan ini untuk mengetahui sifat-sifat serbuk paduan U- 6Zr yang diperoleh dari proses hydriding-dehydriding sebagai kandidat bahan bakar reaktor riset. Serbuk yang diperoleh dari proses hydriding-dehydriding dikenai pengujian, diantaranya pungujian komposisi kimia, densitas, kandungan hidrogen, fasa dan sifat termal. Hasil pengujian komposisi kimia menunjukkan beberapa unsur seperti Al, Ca, Cu, dan Ni melebihi batas yang diijinkan dimana masing-masing unsur terdapat sebesar 202,21 ppm; 214,05 ppm; 61,25 ppm dan 134,53 ppm. Pada pengujian diperolah densitas serbuk U-6Zr sebesar 13,58 g/cm3 dan pada pengujian kandungan hidrogen sisa diperoleh kandungan hidrogen sebesar 0,16 %. Untuk pengujian fasa, diperoleh fasa αU dan δU, sedangkan pada pengujian sifat termal yakni transformasi temperatur terdapat dua puncak yakni puncak pertama terjadi pada temperatur 274 hingga 311 oC dan puncak kedua terjadi pada temperatur 493 hingga 527oC. Puncak pertama terjadi reaksi endotermik dengan menyerap panas sebesar ∆H = 6,23 cal/g tetapi tidak terbentuk fasa baru, sedangkan puncak kedua terjadi reaksi eksotermik dengan mengeluarkan panas sebesar ∆H = -9.34 cal/g dan terbentuk fasa αZr. Sementara itu, dari pengujian kapasitas panas pada temperatur 34 hingga 75 oC, terjadinya penurunan nilai kapasitas panas yang disertai dengan penyerapan panas. Pada temperatur yang lebih tinggi hingga temperatur 437oC nilai kapasitas panas menjadi lebih kecil disertai pengeluaran panas. Reaksi termokimia antara Zr dengan hidrogen sisa menunjukkan terbentuknya fasa αZr yang diindikasikan oleh reaksi

  18. PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWe GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5

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    Andi Sofrany Ekariansyah

    2015-04-01

    Full Text Available Reaktor daya PWR AP1000 yang didesain oleh Westinghouse adalah reaktor Generasi III+ pertama yang telah menerima persetujuan desain dari U.S. Nuclear Regulatory Commission (NRC. Saat ini utilitas China telah memulai pembangunan beberapa unit AP1000 di dua tapak terpilih untuk rencana operasi pada 2013-2015. AP1000 sebagai desain PWR berdasarkan teknologi teruji dari desain PWR lainnya yang dibuat oleh Westinghouse dengan penguatan pada sistem keselamatan pasif dengan demikian dapat dipertimbangkan untuk dibangun di Indonesia bila mengacu pada persyaratan pada PP 43/2006 mengenai Perijinan Reaktor Nuklir. Namun demikian, desain tersebut perlu diverifikasi oleh Technical Support Organization (TSO independen sebelum dapat dibangun di Indonesia. Verifikasi dapat dilakukan menggunakan paket program RELAP5 dalam bentuk analisis kecelakaan. Selama ini analisis kecelakaan PLTN dilakukan untuk tipe PWR 1000 MWe dari generasi II atau tipe konvensional. Mengingat saat ini referensi yang menggambarkan teknologi AP1000 yang menyertakan teknologi keselamatan pasif sudah tersedia maka dilakukan kegiatan pemodelan yang nantinya dapat digunakan untuk melakukan analisis kecelakaan. Metode pengembangan model mengacu pada pedoman IAEA yang terdiri dari pengumpulan data instalasi, pengembangan engineering data dan penyusunan input deck, verifikasi dan validasi data input. Model yang berhasil dikembangkan secara umum telah mewakili sistem AP1000 secara keseluruhan dan dianggap sebagai model dasar. Model tersebut telah diverifikasi dan divalidasi dengan data desain yang terdapat pada referensi dimana respon parameter termohidraulika menunjukkan perbedaan hasil ± 3% selain untuk parameter penurunan tekanan teras yang lebih rendah 13%. Sebagai model dasar, input deck yang diperoleh dapat dikembangkan lebih lanjut dengan mengintegrasikan pemodelan sistem keselamatan, sistem proteksi, dan sistem kendali yang spesifik AP1000 untuk keperluan simulasi keselamatan yang lebih

  19. UJI COBA TEKNOLOGI BIOFILM KONSORSIUM BAKTERI PADA REAKTOR SEMIANAEROB-AEROB UNTUK PENGOLAHAN AIR LIMBAH DI INDUSTRI PENCELUPAN TEKSTIL SKALA RUMAH TANGGA

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    Dewa Ketut Sastrawidana

    2013-04-01

    Full Text Available Penelitian ini bertujuan untuk menganalisis efektifitas teknologi biofilm konsorsium bakteri pada  reaktor semianaerob-aerob ntuk mengolah air limbah pencelupan tekstil. Bakteri pada reaktor semianaerob terdiri dari  Aeromonas sp. Pseudomonas sp, dan Flavobacterium sp. sedangkan pada reaktor aerob terdiri dari Vibrio sp. Plesiomonas sp. dan Enterobacter sp. Perombakan proses pertumbuhan terlekat diawali dengan menumbuhkan konsorsium bakteri pada masing-masing reaktor selama 10  hari menggunakan pada batu vulkanik merah sebagai media pelekatan bakteri. Setelah terbentuk biofilm,selanjutnya digunakan untuk merombak limbah denagn waktu tinggal limbah 2 hari. Hasil penelitian menunjukkan teknologi biofilm cukup efektif diaplikasikan pada skala lapang menghasilkan efisiensi perombakan TSS, BOD dan COD secara berturut-turut sebesar 84,7%; 80,56% dan 90,40%. Uji toksisitas air limbah tekstil menggunakan ikan nila dengan waktu paparan 3 hari menunjukkan bahwa air limbah tekstil sebelum diolah berkatagori toksik ringan dengan nilai EC50 adalah 88,80% sedangkan setelah diolah dalam reaktor biofilm konsorsium bakteri sistem anaerob-aerob selama 2 hari menjadi katagori tidak toksik dengan nilai EC50 sebesar 101,64%. Dengan demikian, pengolahan limbah tektil dengan sistem kombinasi anaerob-aerob menghasilkan kualitas limbah dengan kriteria sudah memenuhi baku mutu untuk dibuang ke lingkungan.

  20. Research nuclear reactor RA - Annual Report 1994; Istrazivacki nuklearni reaktor RA - Izvestaj za 1994. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1994-12-01

    pouzdan rad ovog reaktora. Poslednji, i ujedno najveci zahvat, koji se odnosi na zamenu celokupne instrumentacije je u toku, ali njegova realizacija u 1994. godini kasni zbog zastoja u isporuci opreme koja se izradjuje u Sovjetskom savezu. Izradu ove opreme finansira Medjunarodna agencija za atomsku energiju kroz ugovor sklopljen decembra 1988. godine sa moskovskom firmom Atomenergoexport. Prema tom ugovoru trebalo je da celokupna nova instrumentacija za reaktor RA bude isporucena Institutu u Vinci do kraja 1990. godine, ali je do septembra 1991. godine isporuceno svega 56% od predvidjene kolicine. Od tada je svaka isporuka obustavljena, a razlog je privremena zabrana na sve isporuke opreme za Jugoslaviju izrecena od strane ove organizacije Ujedinjenih nacija. U 1991. godini na reaktoru RA je demontirana postojeca instrumentacija u maksimalno mogucem obimu, kako bi se zadrzala neka osnovna merenja neophodna i u uslovima kada reaktor nije u pogonu. U 1994. godini nastavljen je rad na razradi i dopuni sovjetskog projekta. Kontrola i odrzavanje celopkupne opreme postrojenja, kao i remontni radovi izvrsavani su redovno i efikasno. Veoma obiman remont sekundarnog kola hladjenja reaktora privodi se kraju i bice okoncan u prvoj polovini 1995. a u skladu postojecim zakonskim propisima i sa preporukama MAAE. Kontrola goriva od strane inspektora MAAE obavljana je jedanput mesecno. Na reaktoru RA u 1994. godini radilo je prosecno 47 radnika, sto je dovoljan broj u uslovima remontnih i investicionih radova.

  1. Research nuclear reactor RA - Annual Report 1996; Istrazivacki nuklearni reaktor RA - Izvestaj za 1996. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1996-12-01

    ugovoru trebalo je da celokupna nova instrumentacija za reaktor RA bude isporucena Institutu u Vinci do kraja 1990. godine, ali je do septembra 1991. godine isporuceno svega 80% od predvidjene kolicine. Od tada je svaka isporuka obustavljena, a razlog je privremena zabrana na sve isporuke opreme za Jugoslaviju izrecena od strane ove organizacije Ujedinjenih nacija. Demontirana je postojeca instrumentacija u maksimalno mogucem obimu, kako bi se zadrzala neka osnovna merenja neophodna i u uslovima kada reaktor nije u pogonu. Kontrola i odrzavanje celopkupne opreme postrojenja, kao i remontni radovi izvrsavani su redovno i efikasno. Kontrola goriva od strane inspektora MAAE obavljana je jedanput mesecno. U ovom izvestajnom periodu na reaktoru RA bilo je zaposleno prosecno 43 radnika, sto je dovoljan broj u uslovima remontnih i investicionih radova. Nedostatak finansijskih sredstava za odrzavanje reaktora RA je neresen problem i u ovom periodu. Ovaj izvestaj podeljen je u dve celine: pogon i odrzavanje reaktora i zastita od zracenja na reaktoru RA.

  2. VERIFIKASI PAKET PROGRAM MVP-II DAN SRAC2006 PADA KASUS TERAS REAKTOR VERA BENCHMARK.

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    Jati Susilo

    2015-03-01

    Full Text Available Dalam penelitian ini dilakukan verifikasi perhitungan benchmark VERA pada kasus Zero Power Physical Test (ZPPT teras reaktor Watts Bar 1. Reaktor tersebut merupakan jenis PWR kelas 1000 MWe yang didesain oleh Westinghouse, tersusun dari 193 perangkat bahan bakar 17×17 dengan 3 jenis pengkayaan UO2 yaitu 2,1wt%, 2,619wt% dan 3,1wt%. Perhitungan nilai k-eff dan distribusi faktor daya dilakukan pada siklus operasi pertama teras dengan kondisi beginning of cycle (BOC dan hot zero power (HZP. Posisi batang kendali dibedakan menjadi uncontrolled (semua batang kendali berada di luar teras, dan controlled (batang kendali Bank D didalam teras. Paket program komputer yang digunakan dalam perhitungan adalah MVP-II dan SRAC2006 modul CITATION dengan data pustaka tampang lintang ENDF/B-VII.0. Hasil perhitungan menunjukkan bahwa perbedaan nilai k-eff teras pada kondisi controlled dan uncontrolled antara referensi dengan MVP-II (-0,07% dan -0,014% dan SRAC2006 (0,92% dan 0,99% sangat kecil atau masih dibawah 1%. Perbedaan faktor daya maksimum teras pada kondisi controlled dan uncontrolled dengan referensi dengan MVP-II adalah 0,38% dan 1,53%, sedangkan dengan SRAC2006 adalah 1,13% dan -2,45%. Dapat dikatakan bahwa kedua paket program komputer menunjukkan hasil perhitungan yang sesuai dengan nilai referensi. Dalam hal penentuan kekritisan teras, maka hasil perhitungan MVP-II lebih konservatif dibandingkan dengan SRAC2006. Kata kunci : MVP-II, SRAC2006, PWR, VERA   In this research, verification calculation for VERA core physics benchmark on the Zero Power Physical Test (ZPPT of the nuclear reactor Watts Bar 1. The reactor is a 1000 MWe class of PWR designed by Westinghouse, arranged from 193 unit of 17×17 fuel assembly consisting 3 type enrichment of UO2 that are 2.1wt%, 2.619wt% and 3.1wt%. Core power factor distribution and k-eff calculation has been done for the first cycle operation of the core at beginning of cycle (BOC and hot zero power (HZP. In this

  3. Research nuclear reactor RA - Annual Report 1991; Istrazivacki nuklearni reaktor RA - Izvestaj za 1991. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1992-01-01

    zahvat, koji se odnosi na zamenu celokupne instrumentacije je u toku, ali njegova realizacija u 1991. godini kasni zbog zastoja u isporuci opreme koja se izradjuje u Sovjetskom savezu. Izradu ove opreme finansira Medjunarodna agencija za atomsku energiju kroz ugovor sklopljen decembra 1988. godine sa moskovskom firmom Atomenergoexport. Prema tom ugovoru trebalo je da celokupna nova instrumentacija za reaktor RA bude isporucena Institutu u Vinci do kraja 1990. godine, ali je do septembra 1991. godine isporuceno svega 56% od predvidjene kolicine. Od tada je svaka isporuka obustavljena, a razlog je privremena zabrana na sve isporuke opreme za Jugoslaviju izrecena od strane ove organizacije Ujedinjenih nacija. U 1991. godini na reaktoru RA je demontirana postojeca instrumentacija u maksimalno mogucem obimu, kako bi se zadrzala neka osnovna merenja neophodna i u uslovima kada reaktor nije u pogonu. Pri kraju je izrada odredjenih konstrukcionih elemenata za novu instrumentaciju, koju je prihvatio da realizuje Institut u Vinci, kako bi se ubrzala realizacija ovog projekta. Ako sva predvijena oprema ne bude isporucena do kraja marta 1992. godine, nece moci da se otpocne sa probnim radom reaktora u prvoj polovini 1993. godine, kako je planirano. Na realizaciji projekta u 1991. godini ucestvovalo je efektivno 53 radnika, sto je dovoljan broj u uslovima remontnih i investicionih radova. Godisnji izvestaj o radu nuklearnog reaktora RA za 1992. godinu sastoji se od dva dela: prvi deo obuhvata pogon i odzavanje reaktora RA, a drugi poslove zastite od zracenja na reaktoru RA.

  4. RA Research reactor, Annual report 1988; Istrazivacki nuklearni reaktor RA - Izvestaj za 1988. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1988-12-15

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1989. godinu, koji finansira republicka zajednica za nauku SR Srbije po ugovoru br. 3705/1 sastoji se iz dva dela. Prvi deo obuhvata pogon i odrzavanje nuklearnog reaktora RA, sto predstavlja obavezu Odeljenja za reaktorski inzenjering u sastavu OOUR Instituta za multidisciplinarna istrazivanja i inzenjering RA. Drugi deo obuhvata poslove zastite od zracenja na reaktoru RA, sto predstavlja obavezu OOUR Instituta za zastitu od zracenja 'Zastita'. Naucno vece Instituta za multidisciplinarna istrazivanja i inzenjering RA ocenilo je da sadrzina ovog izvestaja odgovara izvrsenim poslovima na reaktoru RA u 1989. godini. Pozitivno se ocenjuje obim i kvalitet izvrsenih radova kako u pogledu odrzavanja i rekonstrukcije reaktora, tako i u pogledu poslova zastite od zracenja izvrsenih kod njega.

  5. Research nuclear reactor RA - Annual Report 1989; Istrazivacki nuklearni reaktor RA - Izvestaj za 1989. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1989-12-15

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1989. godinu, koji finansira republicka zajednica za nauku SR Srbije po ugovoru br. 3705/1 sastoji se iz dva dela. Prvi deo obuhvata pogon i odrzavanje nuklearnog reaktora RA, sto predstavlja obavezu Odeljenja za reaktorski inzenjering u sastavu OOUR Instituta za multidisciplinarna istrazivanja i inzenjering RA. Drugi deo obuhvata poslove zastite od zracenja na reaktoru RA, sto predstavlja obavezu OOUR Instituta za zastitu od zracenja 'Zastita'. Naucno vece Instituta za multidisciplinarna istrazivanja i inzenjering RA ocenilo je da sadrzina ovog izvestaja odgovara izvrsenim poslovima na reaktoru RA u 1989. godini. Pozitivno se ocenjuje obim i kvalitet izvrsenih radova kako u pogledu odrzavanja i rekonstrukcije reaktora, tako i u pogledu poslova zastite od zracenja izvrsenih kod njega.

  6. Research nuclear reactor RA - Annual Report 1997; Istrazivacki nuklearni reaktor RA - Izvestaj za 1997. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Vinca Institute of Nuclear Sciences, Beograd (Serbia and Montenegro)

    1997-12-01

    RA reactor is not in operation since 1984, activities related to revitalisation of the RA reactor started in 1986. The planned actions related to renewal of the reactor components were finished except for the most important action, related to exchange of complete reactor instrumentation which was delayed. Only 80% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the sanctions imposed to our country. The existing RA reactor instrumentation was dismantled. Control and maintenance of the reactor components was done regularly and efficiently. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. There have been on the average 42 employees at the RA reactor which is considered sufficient for maintenance and repair conditions. The problem of financing the reactor activities and maintenance remains unsolved. Research reactor RA Annual report for year 1997 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. [Serbo-Croat] Reaktor RA nije u pogonu od 1984, aktivnosti na revitalizaciji, rekostrukciji i modernizaciji reaktorski sistema zapocete su 1986. godine. Okoncan je niz zahvata na opremi postrojenja kojima ce se u narednom periodu omoguciti kontinualan i pouzdan rad ovog reaktora. Poslednji, i ujedno najveci zahvat, koji se odnosi na zamenu celokupne instrumentacije kasni zbog zastoja u isporuci opreme koja se izradjuje u Sovjetskom savezu. Do septembra 1991. godine isporuceno svega 80% od predvidjene kolicine. Od tada je svaka isporuka obustavljena, a razlog je privremena zabrana na sve isporuke opreme za Jugoslaviju usled sankcija uvedenih od strane organizacije Ujedinjenih nacija. Demontirana je postojeca instrumentacija. Kontrola i odrzavanje celopkupne opreme postrojenja, kao i remontni radovi izvrsavani su redovno i efikasno. Kontrola goriva od strane safeguard inspektora MAAE obavljana

  7. RA Research reactor, Annual report 1970 - Operation and maintenance; Istrazivacki nuklearni reaktor RA - Izvestaj za 1970. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1970-12-15

    the pool. Refuelling was done three times by 'mixing' the irradiated fuel slugs with 5 to 6 fresh ones. Total exposure of the staff was increased due to activities during the shut-down period. Individual maximum exposure dose was not higher than 3.5 R. Cooperation with the French partner, concerned with the possibility of using 93% enriched fuel instead of the present low enriched, was continued. This would enable achievement of higher neutron flux, order of magnitude of 10{sup 14} n/cm{sup 2} sec. It is mentioned that there is doubt whether the plan for 1971 could be fulfilled for financial difficulties. [Serbo-Croat] Reaktor RA je u 1970. godini radio na nominalnoj snazi 160 dana i 40 dana na manjim snagama. Ukupni rad iznosio je 25968 MWh odnosno 3.87% vise od planiranog. Plan rada razlikovao se od prethodnih godina zbog slanja teske vode na rekoncentraciju u Francusku. Izotopski sastav teske vode je opao na 99.05% a posle rekoncentracije iznosi 99,96%. Odstupanje od plana rada u septembru mesecu bilo je izazvano kasnjenjem prispeca teske vode usled administrativnih teskoca i transporta. Pocetak kampanje odlozen je i zbog posledica ostecenja kosuljice jednog gorivnog elementa, sto se dogodilo odmah po pocetku rada pa je reaktor bio zaustavljen. U oktobru i novembru reaktor je radio 28 odnosno 25 dana respektivno sto je omogucilo da se nadoknadi izgubljeno vreme. Reaktor je koriscen za ozracivanja i eksperimente za 390 korisnika od cega 340 iz Instituta i 50 za korisnika izvan Instituta. Ovaj izvestaj sadrzi detaljne podatke o radu i eksperimentima koji su obavljani. Zakljucuje se da je reaktor radio uspesno prema planu rada. Kracih zastoja u radu bilo je samo zbog teskoca sa cevovodima tehnicke vode kao posledica klizanja zemljista. Reaktor je samo dva puta sigurnosno zaustavljan zbog neispravnosti opreme odnosno laznog signala elektronske aparature sistema upravljanja. Vreme dok reaktor nije radio iskorisceno je za inspekciju unutrasnjosti reaktorskog suda. Pomocu

  8. PERFORMA OKSIDASI METAN PADA REAKTOR KONTINYU DENGAN PENINGKATAN KETEBALAN LAPISAN BIOCOVER LANDFILL

    Directory of Open Access Journals (Sweden)

    Opy Kurniasari

    2013-11-01

    methane through the form of methanol metabolite. ABSTRAKPenanganan sampah kota di Indonesia pada umumnya dilakukan pada tempat pemrosesan akhir sampah (TPA, yang sebagian besar dilakukan dengan cara pengurugan (landfilling yang cenderung bersifat anaerob (tidak ada oksigen. Cara pengurugan ini biasanya dioperasikan lapis perlapis sehingga memungkinkan terjadinya proses anaerob. Pada kondisi ini dipastikan biogas, yaitu gas metana (CH4 dan CO2, akan muncul. Metana adalah gas rumah kaca dengan potensi pemanasan global lebih besar dari CO2, dan dapat mengabsorpsi radiasi infra merah 23 kali lebih efisien dari CO2 pada periode lebih dari 100 tahun. Salah satu cara yang dapat dilakukan untuk mengurangi gas metana dari landfill yang lepas ke alam adalah dengan mengoksidasinya dengan memanfaatkan material penutup landfill (biocover sebagai media mikroorganisma pengoksidasi metana. Aplikasi kompos sebagai material penutup landfill merupakan pendekatan dengan biaya rendah untuk mereduksi emisi gas dari landfill sehingga cocok untuk negara berkembang. Biocover yang digunakan pada penelitian ini adalah kompos landfill mining, yaitu kompos yang terdegradasi secara alami di landfill. Tujuan penelitian ini adalah mengevaluasi kemampuan biocover kompos landfill mining dalam mengoksidasi metana pada ketebalan lapisan tertentu dengan kondisi aliran kontinyu. Tiga buah reaktor kolom yang digunakan terbuat dari flexy glass berukuran tinggi 70 cm dan diameter 15 cm. Gas metana dialirkan dari bawah reaktor secara kontinyu dengan laju alir 5 ml/menit. Kolom diisi dengan biocover kompos landfill mining dengan ketebalan lapisan 5, 25, 35 dan 60 cm. Hasil percobaan menunjukkan bahwa semakin tebal lapisan biocover, semakin tinggi efisiensi oksidasi metana. Efisiensi oksidasi yang diperoleh pada setiap ketebalan lapisan 15, 25, 35 dan 60 cm adalah masing-masing 56,43%, 63,69%, 74,58% dan 80,03%, dengan laju oksidasi 0,287 mol m-2 d-1 dan fraksi oksidasi 97%. Hasil oksidasi yang diperoleh tersebut

  9. ANALISIS PENGARUH IRADIASI FLUENS NEUTRON CEPAT TERHADAP BERILIUM REFLEKTOR REAKTOR RSG-GAS

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    Sri Kuntjoro

    2015-04-01

    Full Text Available Telah dilakukan analisis iradiasi fluens neutron cepat terhadap berilium reflektor reaktor RSG-GAS. Analisis dilakukan dengan cara melakukan pengukuran fluks neutron di posisi berilium elemen dan berilium blok yang berfungsi sebagai reflector. Selanjutnya dilakukan perhitungan untuk menentukan apakah ada pengaruh fluens neutron selama berilium berada di teras reaktor. Selain cara tersebut dilakukan pula visualisasi untuk memastikan ada tidaknya deformasi pada berilium akibat iradiasi. Hasil pengukuran fluks dan fluens neutron cepat maksimal pada daya 200 kW untuk berilium elemen posisi E-2 sebesar 2,30E+07 n/cm2s dan 4,19E+17 n/cm2, J-8 sebesar 3,70E+07 n/cm2s dan 6,74E+17 n/cm2. Hasil pengukutan pada posisi B-3 sebesar 2,19E+12 n/cm2s dan 3,99E+22 n/cm2, G-10 sebesar 2,12E+12 n/cm2s dan 3,86E+22 n/cm2, serta berilium blok posisi (5-6 sebesar 5,02E+07 n/cm2s dan 9,15E+17 n/cm2, (C-D sebesar 2,32E+07 n/cm2s dan 4,23E+17 n/cm2. Deformasi yang diperoleh untuk berilium elemen (∆L/L posisi E-2 sebesar 1,12E-08, J-8 sebesar 1,84E-08, B-3 sebesar 1,60E-03, posisi G-10 sebesar 1,55E-03, sedangkan pada berilium blok di posisi 5-6 sebesar 2,52E-08 dan C-D sebesar 1,13E-08. Dari hasil ini disimpulkan tidak terjadi deformasi pada berilium elemen dan berilium blok. Hasil ini dibuktikan pula dari pengamatan visual, dimana tidak terlihat adanya deformasi pada berilium tersebut. Kata kunci : fluks, fluens, berilium elemen, berilium blok   Analysis of influence fast neutron fluence irradiated to the RSG-GAS beryllium reflector have been done. Methods of analysis was carried out by measuring fluxs neutron in beryllium element and block positio that function as reflector. The calculation done for determination it is there any influence of neutron as long as beryllium in the core. Bisede that, visualization done to make sure it there is any deformation at beryllium as efect of irradiation. Fluxs and fluences of beryllium element measurement result in 200 k

  10. RA Research reactor, Annual report 1971; Istrazivacki nuklearni reaktor RA - Izvestaj za 1971. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1971-12-15

    effects. In the introduction of this report it has been emphasised that the decision makers should have in mind the negative effects of low budget on the reactor safe and reliable operation. For the sake of reactor, decision about the future operation and financing should be done as soon as possible, either to cease operation or continue with adequate financial support. [Serbo-Croat] Reaktor RA je u 1971. godini radio na nominalnoj snazi 190 dana i 50 dana na manjim snagama. Ukupni rad iznosio je 31606 MWh odnosno 5,3% vise od planiranog, sto je najvisa vrednost od kako je reaktor pusten u rad. Reaktor je koriscen za ozracivanja i eksperimente za 425 korisnika od cega 370 iz Instituta i 55 za korisnika izvan Instituta. Ovaj izvestaj sadrzi detaljne podatke o radu i eksperimentima koji su obavljani. Odstupanja od plana, odnosno veceg ostvarenog rada bilo je u junu i decembru usled posebnih zahteva korisnika. Ukupni broj prekida rada bio je manji od svih prethodnih godina, uglavnom zbog manjeg broja nestanka napona u vreme rada reaktora. U toku godine bilo je samo jedno sigurnosno zaustavljanje, ciji je uzrok bila pojava laznog signala opreme za zastitu reaktora. Nijednog duzeg prekida rada nije bilo zbog neispravnosti opreme. Kracih prekida bilo je usled kidanja spojki na potisnom cevovodu tehnicke vode, sto je bilo izazvano klizanjem zemljista u podrucju crpne stanice na Dunavu. Ukupna doza ozracivanja ljudstva bila je manja nego prethodnih godina. Nije bilo ni jednog akcidenta niti slucaja koji bi se mogao nazvati akcidentom. Dekontaminirano je znatno manje povrsina nego ranijih godina. Zakljuceno je da je uspesan rad reaktora u 1971. godini rezultat valjanog rada u prethodnim godinama. Medjutim usled jos nedefinisane politike u pogledu buduceg rada, odnosno neizvesnosti u vezi finansiranja, neki poslovi su obustavljeni. Tu spada proucavanje mogucnosti prelaska na koriscenje visokoobogacenog goriva sto bi povecalo korisni neutronski fluks i ucinilo reakor konkurentnim za

  11. PEMODELAN KOLIMATOR DI RADIAL BEAM PORT REAKTOR KARTINI UNTUK BORON NEUTRON CAPTURE THERAPY

    Directory of Open Access Journals (Sweden)

    Bemby Yulio Vallenry

    2015-03-01

    Full Text Available Salah satu metode terapi kanker adalah Boron Neutron Capture Therapy (BNCT. BNCT memanfaatkan tangkapan neutron oleh 10B yang terendapkan pada sel kanker. Keunggulan BNCT dibandingkan dengan terapi radiasi lainnya adalah tingkat selektivitas yang tinggi karena tingkatannya adalah sel. Pada penelitian ini dilakukan pemodelan kolimator di radial beamport reaktor Kartini sebagai dasar pemilihan material dan manufature kolimator sebagai sumber neutron untuk BNCT. Pemodelan ini dilakukan dengan simulasi menggunakan perangkat lunak Monte Carlo N-Particle versi 5 (MCNP 5. MCNP 5 adalah suatu paket program untuk memodelkan sekaligus menghitung masalah transpor partikel dengan mengikuti sejarah hidup neutron semenjak lahir, bertranspor pada bahan hingga akhirnya hilang karena mengalami reaksi penyerapan atau keluar dari sistem. Pemodelan ini menggunakan variasi material dan ukurannya agar menghasilkan nilai dari tiap parameter-parameter yang sesuai dengan rekomendasi I International Atomic Energy Agency (IAEA untuk BNCT, yaitu fluks neutron epitermal (Фepi > 9 n.cm-2.s-1, rasio antara laju dosis neutron cepat dan fluks neutron epitermal (Ḋf/Фepi 0,7. Berdasarkan hasil optimasi dari pemodelan ini, material dan ukuran penyusun kolimator yang didapatkan yaitu 0,75 cm Ni sebagai dinding kolimator, 22 cm Al sebagai moderator dan 4,5 cm Bi sebagai perisai gamma. Keluaran berkas radiasi yang dihasilkan dari pemodelan kolimator radial beamport yaitu Фepi = 5,25 x 106 n.cm-2s-1, Ḋf/Фepi =1,17 x 10-13 Gy.cm2.n-1, Ḋγ/Фepi = 1,70 x 10-12 Gy.cm2.n-1, Фth/Фepi = 1,51 dan J/Фepi = 0,731. Berdasarkan penelitian ini, hasil optimasi 5 parameter sebagai persyaratan kolimator untuk BNCT yang keluar dari radial beam port tidak sepenuhnya memenuhi kriteria yang direkomendasikan oleh IAEA sehingga perlu dilakukan penelitian lebih lanjut agar tercapainya persyaratan IAEA. Kata kunci: BNCT, radial beamport, MCNP 5, kolimator   One of the cancer therapy methods is

  12. RA Research reactor, Annual report 1968 - Operation and maintenance; Istrazivacki nuklearni reaktor RA - Izvestaj za 1968. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1968-12-15

    During 1968, the RA Reactor was operated at nominal power of 6.5 MW for 190 days, and during 50 days at lower power levels. Total production amounted to 31051 MWh which is 3.5% higher than planned. reactor was used for irradiation and experiments according to the demand of 600 users, of which 517 from the Institute and 83 externals users. This report contains detailed data about reactor power and experiments performed in 1968. It is concluded that the reactor operation was more successful than during previous years. There was only one longer interruption which lasted 27 hours because of the power cut on the cable for the pump station on Danube. Number of safety shutdowns were at the same level as during last year. The only significant incident in 1968 was air contamination with the radioactive argon in the reactor hall. The reactor operation was not interrupted although the hall was evacuated for two hours. The was no significant exposure of the staff. In April and September the integral dosed were higher than during other months because of the accident during refueling (mixing the slugs with irradiated and fresh fuel). There was no significant surface contamination, i.e. the decontaminated surface were negligible. Due to 'mixing' refueling scheme. [Serbo-Croat] Reaktor RA je u 1968. godini radio na nominalnoj snazi od 6,5 MW 190 dana i 50 dana na manjim snagama. Ukupni rad iznosio je 31051 MWh odnosno 3,5% vise od planiranog. Reaktor je koriscen za ozracivanja i eksperimente za 600 korisnika od cega 517 iz Instituta i 83 za korisnika izvan Instituta. Ovaj izvestaj sadrzi detaljne podatke o radu i eksperimentima koji su obavljani. Zakljucuje se da je reaktor radio uspesnije nego prethodnih godina. U toku 1968. godine samo je jedan duzi prekid u radu od 27 casova izazvan zbog proboja kablovske glave na odvodu za pumpnu stanicu na Dunavu. Sigurnosna zaustavljanja bila su na proslogodisnjem nivou. Jedini znacajniji incident u 1968. godini, bio je kontaminacija vazduha

  13. Pengaruh Laju Alir Inlet Reaktor MSL terhadap Reduksi BOD, COD, TSS, dan Minyak/Lemak Limbah Cair Industri Minyak Goreng

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    Salmariza Sy

    2017-06-01

    Full Text Available This research was conducted by treating edible oil industry wastewater used Multi Soil Layering (MSL method. The MSL reactor was built from a 200x120x200 cm concrete basin. Andisol soil was mixed with sawdust and fine charcoal at each ratio 5:1:1 based on dry weight as an impermeable layer. The flow rate variations were 250, 500, 1000, and 1500 L/m2.day. The observed pollutant parameters were BOD, COD, TSS, oil/fat, and pH. The results showed that MSL reactor was effective to decrease the pollutant content of edible oil industry wastewater. The reactor could reduce concentration of effluent parameters below standard except for oil/fat parameters at high flow rates. In the effluent was found BOD 0.66-14.22 mg/L, COD 5-69 mg/L, TSS 9-26 mg/L, and oil/fat 2-9 mg/L. The flow rate had an effect on reduction efficiency of BOD, COD, TSS, and oil/fat but did not effect pH as all flow rate could raise pH 6.37-6.95 became pH 6.99-7.24. The lower the flow rate the higher the reduction efficiency. The reduction efficiency at flow rates 250 and 1500 L/m2 days for BOD were 99% and 86%, COD were 96% and 71%, TSS were 88% and 77%, and oil/fat were 80% and 60%.ABSTRAK  Penelitian ini dilakukan dengan mengolah air limbah industri minyak goreng menggunakan metoda Multi Soil Layering (MSL. Reaktor MSL dibuat dari beton berbentuk bak ukuran 200x120x200 cm. Tanah andisol dicampur dengan serbuk gergaji dan arang halus pada rasio masing-masing 5:1:1 berdasarkan berat kering sebagai penyusun lapisan impermeable. Variasi laju alir yaitu 250, 500, 1000, dan 1500 L/m2.hari. Parameter pencemar yang dianalisis meliputi BOD, COD, TSS, minyak/lemak, dan pH. Hasil penelitian menunjukkan bahwa reaktor MSL sangat efektif untuk menurunkan kandungan zat pencemar limbah cair industri minyak goreng. Reaktor dapat mereduksi konsentrasi parameter outlet sampai dibawah baku mutu yang distandarkan kecuali untuk parameter miyak/lemak pada perlakuan laju alir tinggi. Pada effluen

  14. RA Research reactor, Annual report 1969; Istrazivacki nuklearni reaktor RA - Izvestaj za 1969. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1969-12-15

    years of operation as well as purification of heavy water. [Serbo-Croat] Reaktor RA je u 1969. godini radio na nominalnoj snazi 200 dana i 15 dana na manjim snagama. Ukupni rad iznosio je 31131 MWh odnosno 3.77% vise od planiranog. Reaktor je koriscen za ozracivanja i eksperimenta prema zahtevima 463 korisnika iz Instituta i 63 izvan Instituta. Ovaj izvestak sadrzi detaljne podatke o snazi na kojoj je reaktor radio tokom 1969. godine i o uradjenim eksperimentima. Zakljuceno je da je reaktor RA uspesno radio prema planu rada. Da nije bilo problema sa napajanjem elektricnom energijom tokom poslednja tri meseca i niskog vodostaja Dunava u septembru i oktobru protekle godine bila bi to najuspesnija godina od pustanja reaktora u pogon. Broj sigurnosnih zaustavljanja nije bio veci u odnosu na prethodne dve godine i pored poteskoca u poslednjem kvartalu. Osoblje je bilo izlozeno povecanim dozama zracenja usled tri incidenta. Jedan je bio raspadanje kenera sa srebrom (zbog duzeg stajanja u aktivnoj zoni), sto je uzrokovalo kontaminaciju radne platforme, tako da je fon porastao za 10 do 100 puta od normalnohg Druga dva slucaja su bila otkazivanje uredjaja za mesanje goriva u tehnoloskim kanalima. Zamena goriva je radjena cetiri puta u toku godine, utroseno je 499 svezih gorivnih elemenata. Primenjena je metoda mesanja svezih gorivnih elemenata sa koriscenim gorivnim elementima u gorivnom kanalu. Dekontaminacija povrsina bila je na nivou prethodnih godina i pored problema sa srebrom. Kako su sa reaktora tokom godine otisla dva saradnika sa visokom spremom broj ljudi je opao na neophodan minimum za pogon i odrzavanje reaktora. Navrsavajuci u ovoj godini deset godina rada moze se reci da su rad i stanje opreme na tehnicki solidnom nivou. Kako bi se posle deset godina rada izvrsila kontrola vaznih komponenti reaktora i obavila rekoncentracija teske vode, za 1970. godinu je planirano da se proizvodnja smanji na 25000 MWh, a baziran je na istim principima kao i planovi za prethodne

  15. PENGARUH KONSENTRASI ZrO2 TERHADAP KORELASI PERPINDAHAN PANAS NANOFLUIDA AIR-ZrO2 UNTUK PENDINGIN REAKTOR

    Directory of Open Access Journals (Sweden)

    K.A. Sudjatmi

    2015-03-01

    Full Text Available Sejalan dengan perkembangan konsep keselamatan pasif pada sistem keselamatan PLTN, maka sistem perpindahan panas konveksi alam memegang peranan penting. Pemakaian nanofluid sebagai fluida pendingin pada sistem keselamatan nuklir dapat digunakan pada Sistem Pendingin Teras Darurat dan Sistem Pendingin Pengungkung Luar Reaktor. Beberapa peneliti telah melakukan studi desain konseptual aplikasi nanofluid untuk meningkatkan keselamatan AP1000 dan sistem pendingin teras darurat pada reaktor daya eksperimen. Penerapan nanofluida juga mulai dikembangkan melalui hasil penelitian perpindahan panas konveksi alamiah pada sub-buluh dengan nanofluida sebagai fluida kerjanya sangat dibutuhkan. Penelitian ini bertujuan untuk menentukan pengaruh perubahan konsentrasi ZrO2 terhadap korelasi perpindahan panas konveksi alamiah dengan pendekatan eksperimental. Data eksperimental yang diperoleh digunakan untuk mengembangkan korelasi umum empirik perpindahan panas konveksi alamiah. Metode penelitian dengan menggunakan alat uji sub-buluh vertikal dengan geometri segitiga dan segiempat menggunakan air dan nanofluida air-ZrO2 sebagai fluida kerjanya. Konsentrasi nanopartikel dalam larutan yang digunakan sebesar 0,05 %, 0,10% dan 0,15 % dalam persen berat. Hasil penelitian menunjukan bahwa untuk bilangan Rayleigh yang sama, kemampuan pemindahan kalor oleh nanofluida air-ZrO2 lebih baik dari pada pemindahan kalor oleh air. Namun peningkatan konsentrasi nanofluida tidak selalu mendapatkan kemampuan pemindahan kalor yang lebih baik. Kata kunci: nanofluida air-ZrO2, konveksi alamiah, sub-buluh segitiga, sub-buluh segi segiempat   In line with the development of the passive safety concept for the safety systems of nuclear power plants, the natural convection heat transfer system plays an important role. The nanofluid as coolant fluid on nuclear safety system can be used in Emergency core cooling system and in reactor coolant system confinement. Several researchers have

  16. RA Research reactor, Annual report 1973; Istrazivacki nuklearni reaktor RA - Izvestaj za 1973. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1973-12-15

    During 1973, RA reactor was operated at nominal power for 4687 hours and 54 hours at lower power levels. The total production was 30504 MWh which is 1.6% higher than planned. Practically there was no discrepancies from the plan, since the action plan was corrected at the beginning of this year caused by the demand of changing the schedule for refuelling for the purpose of 'power excursion' experiment. The reactor was used for irradiation and experiments according to the demand of 336 users. This report contains detailed data about reactor power and experiments performed in 1973. Total number safety shutdowns was 12, of which 7 were caused by power cuts. Three shutdowns caused by failures of the equipment were caused by failures of new electronic tubes. Two shutdowns were caused by the operators. There have been three shorter interruptions announced power cuts. Total personnel exposure dose was lower than during previous years. There were no accidents during this year. Decontamination of surfaces was less than during previous years. Practically there was no surface contamination, and the quantity of collected radioactive waste was lower than previously. There were no liquid radioactive effluents. It was concluded that the successful operation in 1973 has a special significance taking into account the financial crisis. There still remains a number of unsolved problems related to: completing the inventory of spare parts, exchange of some elements of the equipment, exchange of instrumentation, and purchase of the highly enriched fuel. [Serbo-Croat] Reaktor RA je u 1973. godini radio na nominalnoj snazi 4687 sati i 54 sata na manjim snagama. Ukupni rad iznosio je 30504 MWh odnosno 1,6% vise od planiranog. Prakticno nije bilo odstupanja od plana rada koji je pocetkom godine korigovan zbog promene planiranih izmena goriva usled izvodjenja eksperimenta 'ekskurzije snage'. Reaktor je koriscen za ozracivanja i eksperimente za 336 korisnika. Ovaj izvestaj sadrzi detaljne

  17. RA Research nuclear reactor - Annual report 1987; Istrazivacki nuklearni reaktor RA - Izvestaj o radu za 1987. godinu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1987-12-15

    Annual report concerning the project 'RA research nuclear reactor' for 1987, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1987. godinu, koji finansira republicka zajednica za nauku SR Srbije po ugovoru br. 3509/1 sastoji se iz dva dela. Prvi deo obuhvata pogon i odrzavanje nuklearnog reaktora RA, sto predstavlja obavezu Odeljenja za reaktorski inzenjering u sastavu OOUR Instituta za multidisciplinarna istrazivanja i inzenjering RA. Drugi deo obuhvata poslove zastite od zracenja na reaktoru RA, sto predstavlja obavezu OOUR Instituta za zastitu od zracenja 'Zastita'. Naucno vece Instituta za multidisciplinarna istrazivanja i inzenjjring RA ocenilo je da sadrzina ovog izvestaja odgovara izvrsenim poslovima na reaktoru RA u 1989. godini. Pozitivno se ocenjuje obim i kvalitet izvrsenih radova kako u pogledu odrzavanja i rekonstrukcije reaktora, tako i u pogledu poslova zastite od zracenja izvrsenih kod njega.

  18. A pulsed fast reactor; Un reacteur pulse a neutrons rapides; Impul'snyj reaktor na bystrykh nejtronakh; Reactor rapido pulsado

    Energy Technology Data Exchange (ETDEWEB)

    Blokhin, G. E.; Blokhintsev, D. I.; Blyumkina, Yu. A.; Bondarenko, I. I.; Deryagin, B. N.; Zajmovskij, A. S.; Zinov' ev, V. P.; Kazachkovskij, O. D.; Krasnoyarov, N. V.; Lejpunskij, A. I.; Malykh, V. A.; Nazarov, P. M.; Nikolaev, S. K.; Stavisskij, Yu. Ya.; Ukraintsev, F. I.; Frank, I. M.; Shapiro, F. Ji.; Yazvitskij, Yu. S. [Akademiya Nauk, Moscow, SSSR (Russian Federation)

    1962-03-15

    los impulsos de potencia. Asimismo, se efectuaron mediciones del periodo de los neutrones instantaneos, de la fraccion efectiva de neutrones retardados y de los coeficientes de variacion de la reactividad en funcion de la temperatura. (author) [Russian] Impul'snyj reaktor na bystrykh nejtronakh (IBR) rabotaet na nominal'noj moshchnosti v Obedinennom institute yadernykh issledovanij s dekabrya 1960 goda. Reaktor ispol'zuetsya v kachestve impul'snogo istochnika nejtronov dlya fizicheskikh ehksperimentov, provodimykh metodom vremeni proleta. Provodyatsya izmereniya polnogo secheniya, secheniya zakhvata dlya promezhutochnykh nejtronov, issledo- vaniya vzaimodejstviya medlennykh nejtronov s tverdym telom i s zhidkost'yu, izmereniya spektrov nejtronov, ustanavlivayushchikhs ya v. razlichnykh sredakh. V doklade opisany osnovy konstruktsii reaktora i rezul'taty ego issledovanij. Osnovnoj rezhim raboty reaktora-rezhim periodicheskikh impul'sov. Impul'sy moshchnosti voznikayut pri bystrom peremeshchenii podvizhnoj chasti aktivnoj zony reaktora cherez ego nepodvizhnuyu zonu. Podvizhnaya chast' aktivnoj zony zakreplena vo vrashchayushchemsya diske i dvizhetsya so skorost'yu-230 m/sek. Chastota impul'sov moshchnosti mozhet izmenyat'sya s pomoshch'yu vspomogatel'noj podvizhnoj zony v diapazone 2,3-88 im/sek. Srednyaya moshchnost' reaktora - 1 kvt. Polushirina impul'sa moshchnosti - 36 mksek. Reaktor snabzhen sistemoj upravleniya i zashchity, obespechivayushchej avtomaticheskoe podderzhanie srednej moshchnosti reaktora i ego bystruyu ostanovku v sluchae narusheniya rezhima. Reaktor snabzhen sistemoj vakuumirovanny kh nektronovodov, ispol'zuemykh v ehksperimentakh po vremeni proleta. Glavnyj nejtronovod imeet dlinu 1000 m. V protsesse puska i fizicheskikh issledovanij reaktora izuchalos' vliyanie peremeshcheniya organov regulirovaniya i podvizhnykh chastej aktivnoj zony na reaktivnost', izmeryalas' dlitel'nost' impul'sa pri razlichnykh rezhimakh raboty reaktora, izuchalis

  19. APLIKASI TEKNIK AAN DI REAKTOR RSG-GAS PADA PENENTUAN UNSUR ESENSIAL DAN TOKSIK DI DALAM IKAN DAN PAKAN IKAN

    Directory of Open Access Journals (Sweden)

    Saeful Yusuf

    2015-03-01

    Full Text Available Pada makalah ini diuraikan tentang aplikasi teknik AAN (Analisis Aktivasi Neutron dalam penentuan konsentrasi unsur-unsur esensial dan cemaran yang terkandung di dalam beberapa spesies ikan dan pakan ikan. Unsur-unsur esensial yang terkandung dalam pakan ikan buatan juga dianalisis untuk mengetahui pengaruhnya terhadap ikan. Penentuan unsur menggunakan teknik AAN dengan metode perbandingan dan metode k0-AAN. Sampel diiradiasi di reaktor RSG-GAS yang memiliki fluks neutron thermal 5 x 1013 n.cm-2.s-1 pada daya 15 MW. Hasil penelitian menunjukkan bahwa 12 unsur di dalam 11 spesies ikan air laut dan air tawar telah ditentukan yaitu As, Br, Cr, Co, Cs, Fe, Hg, K, Na, Rb, Se and Zn. Konsentrasi cemaran As didalam ikan laut sudah melampaui batas maksimum 1 mg/kg, sedangkan konsentrasi cemaran Hg masih dibawah batas maksimum 0,5 mg/kg, baik untuk ikan laut maupun ikan air tawar. Unsur K dan Na merupakan unsur makroesensial sedangkan unsur Cr, Co, Fe, Se and Zn adalah termasuk unsur mikroesensial. Secara umum ditunjukkan bahwa kandungan mineral didalam ikan laut lebih tinggi konsentrasinya dibandingkan ikan air tawar. Br, Cs dan Rb merupakan unsur-unsur non esensial yang teridentifikasi dalam semua ikan yang dianalisis. Penelitian terhadap pakan ikan air tawar menunjukkan bahwa semua unsur yang teridentifikasi juga terdapat di dalam ikan laut dan ikan air tawar. Hal ini menunjukkan bahwa pakan ikan berkontribusi terhadap konsentrasi unsur di dalam ikan air tawar. Kata kunci : Analisis aktivasi neutron, unsur esensial, unsur cemaran, ikan, pakan ikan   This paper reported on the application of NAA (Neutron Activation Analysis Technique in the determination of the concentration of the essential and toxic elements in some species of fish and fish feed. Determination of elements using instrumental NAA technique with comparison and k0-INAA methods. Samples were irradiated in the RSG-GAS which has a thermal neutron flux  5.0E +13 ncm-2s-1. The results

  20. RA Research reactor, Annual report 1972; Istrazivacki nuklearni reaktor RA - Izvestaj za 1972. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1972-12-15

    During 1972, the total production was 31151 MWh which is 3.8% higher than planned. The reactor was used for irradiation and experiments according to the demand of 381 users, of which 340 from the Institute and 41 external users. This report contains detailed data about reactor power and experiments performed in 1972. Discrepancies from the action plan, meaning higher production was achieved due to special demands of the users. Total number of interruptions was lower than during all the previous years, and were caused mainly due to announced power cuts. There was only on scram shutdown during this year caused by a false signal of the reactor control instrumentation. There were no longer interruptions. One shorter interruption (shorter than 24 hours) caused by removal of a UO{sub 2} capsule from the core, placed there for measuring heat transfer. Total personnel exposure dose was lower than during previous years. One accident caused contamination with gases and aerosols containing mainly shot-living isotopes. Decontamination od surfaces was less than during previous years. Practically there was no surface contamination that would demand action of the decontamination team, except for the regular decontamination after refueling. It was concluded that the successful operation in 1972 has a special significance having taking in account the financial crisis caused by the unresolved status of the reactor. It is emphasised, in the plan for the next year that there is an urgent need of making a long-term plan of rector application. It is indispensable to finish preparatory tasks for replacing the fuel with the highly enriched fuel elements by 1974, and building the core emergency cooling system. [Serbo-Croat] Ukupni rad Reaktora RA je u 1972. godini iznosio je 31151 MWh odnosno 3,8% vise od planiranog. Reaktor je koriscen za ozracivanja i eksperimente za 381 korisnika od cega 340 iz Instituta i 41 za korisnike izvan Instituta. Ovaj izvestaj sadrzi detaljne podatke o radu i

  1. RA Reactor; Reaktor RA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-07-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels. [Serbo-Croat] Ovo (prvo) poglavlje sadrzi sledece: Opis reaktora RA; semu organizacije rada i rukovodjenja; prava i duznosti direktora i rukovodioca pogona reaktora, propise o rezimu rada i kretanja u zgradi reaktora, propise o izvodjenju eksperimenata, propise o unosenju uzoraka u eksperimentalne kanale reaktora.

  2. RA Reactor; Reaktor RA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1978-02-15

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation. [Serbo-Croat] Pored osnovnih karakeristika reaktora RA, organizacije rada i finansijskih pokazatelja, razmatra se stanje opreme reaktora nakon 18 godina rada, pitanja dozvole za rad sa 80% obogacenim gorivom, problem skladistenja isluzenog goriva u bazenu zgrade reaktora i potreba za obnavljanjem komponenti opreme, pre svega elektronske.

  3. RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987; Istrazivacki nuklearni reaktor RA Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1987. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1987-12-15

    RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction. [Serbo-Croat] Reaktor RA nije radio usled zabrane Izvsnog veca Skupstine Srbije od 27. avgusta 1984. U cilju povecanja pouzdanosti rada reaktora a da bi se udovoljilo zakonskim propisima sto je uslov za dobijanje stalne dozvole za rad realizovana su tri velika zahvata na reaktoru RA. Ovi zahvati obuhvatili su izgradnju sistema za hladjenje jezgra reaktora u slucaju nuzde, rekonstukciju postojeceg sistema specijalne ventilacije i rekonstrukciju sistema napajanja elektricnom energijom neophodnih potrosaca reaktora RA. Zapoceti su radovi na modernizaciji intrumentacije reaktora RA, projekat je izradjen u sovjetskoj organizaciji Atomenergoeksport, a trebalo bi da se realizuje do kraja 1989. godine. U cilju povecanja prostora za skladistenje ozracenog nuklearnog goriva i njegovog efikasnijeg koriscenja, izradjen je su projekti za rekonstrukciju postojecih uredjaja za rukovanje gorivom, povecanje smestajnog kapaciteta i preciscavanje vode u bazenima za odlezavanje. Realizaija ovih

  4. Operational report, Advantages of gradual introducing of highly enriched fuel into the RA reactor core from economic aspect and users needs; Radni izvestaj, Prednost postupka parcijalnog uvodjenja visokoobogacenog goriva u reaktor RA sa aspekta ekonomicnosti i potreba korisnika

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-10-14

    The possibility of increasing the neutron flux in the RA reactor was considered for a number of years. The possibilities of reactor reconstruction are not realistic and they should be disregarded. The possibility that remains is to achieve higher neutron flux by improving the fueling scheme and above all by introducing highly enriched fuel into the reactor core. Decision to purchase highly enriched fuel was quicker due to the fact that the 2% enriched uranium fuel is not fabricated any more. There are two procedures for exchanging the fuel in the reactor core: a) removal of partially spent 2% enriched fuel and formation of the core with fresh highly enriched fuel; b) gradually introducing the new fuel into the existing RA reactor core according to a special transfer regime. This report includes some comparative analyses of these two procedures from both economic point of view and the needs of users, as well as some technical conditions. These results are in favour of gradual introducing of new fuel into the reactor core. relevant direct savings amount to 3 000 000 dinars. Some of the most important advantages cannot be estimated in this way. This report does not cover the safety analyses results which are presented in a series of other papers. [Serbo-Croat] Vec vise godina razmatra se mogucnost za povecanje neutronskog fluksa u reaktoru RA. Mogucnosti za rekonstrukciju reaktora RA u tom smislu su minimalne i realno ih treba odbaciti. Prema tome preostaje da se povecanje neutronskog fluksa postigne usavrsavanjem seme izmene goriva, a pre svega uvodjenjem goriva sa visokim stepenom obogacenja u reaktor RA. Donosenje odluke o nabavci visokoobogacenog goriva i njegovom uvodjenju u reaktor ubrzano je i cinjenicom da se staro 2% obogaceno uransko gorivo vise ne proizvodi. Postoje dva postupka za prevodjenje reaktora na ovo gorivo: a) Uklanjanjem poluistrosenog 2% obogacenog goriva iz reaktora i formiranjem jezgra iskljucivo od svezeg visokoobogacenog goriva, b

  5. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor; Istrazivacki nuklearni reaktor RA, Deo 1 - Pogon, odrzavanje i eksploatacija reaktora u 1981. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Milosevic, M; Martinc, R; Kozomara-Maic, S; Cupac, S; Radivojevic, J; Stamenkovic, D; Skoric, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1981-12-15

    biggest difficulty was maintenance of reactor instrumentation. During 1981 the reactor was operated safely, there was no accident nor incident that would affect the safety of reactor personnel or the environment. The testing operation will be continued in 1982,and the experience so far shows that the program would be successfully fulfilled on the whole. [Serbo-Croat] Nuklearni reaktor RA prestao je sa radom nakon martovske kampanje 1979. godine usled pojave talozenja oksihidrata aluminijuma na kosuljicama gorivnih elemenata. Odgovarajucim resenjima Sanitarnog inspektorata Republickog sekretarijata za zdravje i socijalnu politiku SR Srbije i generalnog direktora Instituta za nuklearne nauke 'Boris Kidric', Vinca zabranjen je dalji rad reaktora sve dok se ne utvrde uzroci stvaranja oksihidrata aluminijuma i njihovog talozenja, preduzmu mere za njihovo uklanjanje i ne obezbede potrebni uslovi za normalan nastavak rada reaktora. Do kraja 1979. i tokom 1980. godine, nakon niza izvrsenih analiza i utvrdjivanja uzroka koji su doveli do zaustavljanja rada reaktora, izvrsene su sve neophodne pripreme za ponovno pustanje reaktora u rad. Polazeci od cinjenice da na reaktoru RA ne postoji sistem za hladjenje jezgra u slucaju udesa i da ne postoji adekvatan sistem za filtriranje potencijalno zagadjenog vazduha, a saglasno sa novim propisima o pustanju u rad i probnom radu nuklearnih objekata, Sanitarni inspektorat je doneo privremeno resenje kojim se dozvoljava pustanje reaktora u rad, tj. izvodjenje tzv. 'nultog eksperimenta' uz ogranicenje snage na 1% od vrednosti nominalne snage. Na osnovu dobijene dozvole, reaktor RA je ponovo pusten u rad 21. januara 1981. godine, kada je dostignuta kriticnost sa jezgrom sastavljenim iskljucivo od gorivnih elemenata od 80% obogacenog uranijuma. Eksperiment je zavrsen krajem marta, nakon cega je zatrazena dozvola za probni rad na vecim snagama i potom za rad na punoj snazi. Uzimajuci postojece stanje reaktora RA doneto je resenje kojim se

  6. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1983; Istrazivacki nuklearni reaktor RA - Deo I - Pogon, odrzavanje i eksploatacija nuklearnog reaktora RA u 1983. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Kozomara-Maic, S; Cupac, S; Raickovic, N; Radivojevic, J; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1983-12-15

    After regular shutdown in November 1982, inspection of the fuel elements from the RA reactor core which was done from December 1982 - February 1983 has shown that there are deposits of aluminium oxides on the surface of the fuel cladding. After restart The RA reactor was operated at power levels from 1.8 - 2 MW, with 80% enriched uranium dioxide fuel elements. It was found that there was no corrosion of the fuel element cladding and that it was not possible to find the cause of surface deposition on the cladding surfaces without further operation. It was decided to purify the heavy water permanently during operation and to increase the heavy water flow by operating two pumps. This procedure was adopted in order to decrease the possibility of corrosion. The Safety committee of the Institute has approved this procedure for operating the RA reactor in 1983. The core was made of 80% enriched fuel, critical experiments were done until June 1983, and after that the operation was continued at power levels up to 2 MW. [Serbo-Croat] Pregledom nuklearnog goriva iz tehnoloskih kanala reaktora RA koji je izvrsen u periodu decembear 1982-feburuar 1983. godine nakon zaustavljanja reaktora po isteku novembarske kampanje 1982. godine, ustanovljeno je da ponovo dolazi do stvaranja taloga u obliku hidratisanih oksida aluminiuma na kosuljicama gorivnih elemenata. Nakon ponovnog pustanja u rad, reaktor je do novembra 1981. godine neprekidno bio u pogonu na snagama 1,8 - 2 MW. Jezgro je bilo formirano iskljucivo sa od gorivnih elemenata sa 80% obogacenim uran dioksidom. Utvrdjeno je da kosuljica gorivnog elementa nije korodirala, i da se bez nastval rada ne moze utvrditi uzrok pojave taloga na povrsini kosuljice. Da bi se mogucnost korozije aluminjumskih komponenti u primarnom kolu raktora svela na sto manju meru odluceno je da se vrsi neprekidno preciscavanje teske vode i da se istovremeno poveca protok teske vode radom dve pumpe, Komitet za sigurnost Instituta odobrio je ovakav nacin

  7. External irradiation of the personnel operating the reactor RA at Vinca in the period 1963-1966; Spoljasnje ozracivanje osoblja koje opsluzuje reaktor RA u Vinci u periodu 1963-1966

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M M; Minincic, Z [Institut za nuklearne nauke Vinca, Belgrade (Yugoslavia)

    1967-07-01

    The objective of this paper was the analysis of external radiation exposure of the personnel employed at the RA reactor in Vinca during past four years. During 1963 reactor was not operated because of the general repair and maintenance, and it was operated during 1964, 1965 and 1966. The internal irradiation and the estimation of the beta doses on the skin were not analysed in this paper, they should be treated by separate analyses. The evaluation of external irradiation covered only gamma radiation since yhe neutron doses were negligible. The analysis was based on the the irradiation dose data obtained by film and personal dosemeters. Predmet ovog rada je analiza spoljasnjeg ozracivanja osoblja koje je radilo na reaktoru RA u Vinci u poslednje cetiri godine, u toku kojih je reaktor bio u opstem remontu 1963, odnosno u normalnom radu 1964, 1965 i 1966. U ovom radu nismo se upustali u analizu unutrasnjeg ozracivanja, kao ni procenu doza beta zracenja na kozi, sto treba da bude predmet posebnih analiza. Pri proceni spoljasnjeg ozracivanja operisano je samo sa dozama gama zracenja, jer je ozracivanje neutronima zanemarljivo. Celokupna analiza izvrsena je na osnovu podataka o dozama ozracivanja do kojih se doslo ocitavanjem film o penkalo dozimetara.

  8. External irradiation of the personnel operating the reactor RA at Vinca in the period 1963-1966; Spoljasnje ozracivanje osoblja koje opsluzuje reaktor RA u Vinci u periodu od 1963. do 1966. godine

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M M; Minincic, Z [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1968-06-15

    The paper first gives a survey of the characteristic operations performed on the Vinca reactor RA in which most of the personnel become irradiated. Following is a schematic diagram of the irradiations in the period 1963-1966 in which the reactor was in continual operation. The surveys are given for each month and each year separately, while the irradiated personnel are grouped in several characteristic irradiation dose ranges. In this analysis special emphasis is given to a survey of irradiation of the personnel classified according to their profession, i.e. their post. This kind of analysis is indispensable in planning work, proper disposition of the personnel and undertaking special protective measures for reducing the irradiations (author) [Serbo-Croat] U radu je najpre dat pregled mesta na reaktoru RA, sa topografijom zracenja, na kojima radno osoblje pretezno biva ozracivano. Zatim su dati tabelarno i graficki pregledi ozracivanja za protekli period u kome je reaktor neprekidno radio, tj. od 1963. do 1966. god. Pregledi su dati po mesecima za svaku godinu i to tako sto su ozracivana lica razvrstavana u nekoliko karakteristicnih intervala doza ozracivanja. Posebno mesto u ovoj analizi zauzima pregled ozracivanja osoblja razvrstanog po strukama, odnosno radnim mestima. Ovakva analiza je nuzna za planiranje poslova, pravilan raspored radnog osoblja kao i preduzimanje posebnih zastitnih mera u cilju smanjivanja ozracivanja (author)

  9. The Text of the Agreement of 22 July 1977 between Argentina and the Agency for the Application of Safeguards in Connection with a Contract Concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reaktor Brennelement Union Gmbh Hanau (Federal Republic of Germany) for Co-Operation in the Field of Fabrication of Fuel Elements for Peaceful Nuclear Activities

    International Nuclear Information System (INIS)

    1977-01-01

    The text of the Agreement of 22 July 1977 between Argentina and the Agency for the application of safeguards in connection with the Contract of 13 August 1976 concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reaktor Brennelement Union GmbH (Federal Republic of Germany) for co-operation in the field of fabrication of fuel elements for peaceful nuclear activities is reproduced in this document for the information of all Members. The Agreement entered into force, pursuant to Section 26, on 22 July 1977.

  10. The Text of the Agreement of 22 July 1977 between Argentina and the Agency for the Application of Safeguards in Connection with a Contract Concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reaktor Brennelement Union Gmbh Hanau (Federal Republic of Germany) for Co-Operation in the Field of Fabrication of Fuel Elements for Peaceful Nuclear Activities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-11-30

    The text of the Agreement of 22 July 1977 between Argentina and the Agency for the application of safeguards in connection with the Contract of 13 August 1976 concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reaktor Brennelement Union GmbH (Federal Republic of Germany) for co-operation in the field of fabrication of fuel elements for peaceful nuclear activities is reproduced in this document for the information of all Members. The Agreement entered into force, pursuant to Section 26, on 22 July 1977.

  11. Research nuclear reactor RA - Annual Report 1995, with comparative review for period 1991-1995; Istrazivacki nuklearni reaktor RA - Izvestaj za 1995. godinu, uz uporedni pregled za period 1991-1995

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1995-12-01

    firmom Atomenergoexport. Prema tom ugovoru trebalo je da celokupna nova instrumentacija za reaktor RA bude isporucena Institutu u Vinci do kraja 1990. godine, ali je do septembra 1991. godine isporuceno svega 56% od predvidjene kolicine. Od tada je svaka isporuka obustavljena, a razlog je privremena zabrana na sve isporuke opreme za Jugoslaviju izrecena od strane ove organizacije Ujedinjenih nacija. Demontirana je postojeca instrumentacija u maksimalno mogucem obimu, kako bi se zadrzala neka osnovna merenja neophodna i u uslovima kada reaktor nije u pogonu. Kontrola i odrzavanje celopkupne opreme postrojenja, kao i remontni radovi izvrsavani su redovno i efikasno. Kontrola goriva od strane inspektora MAAE obavljana je jedanput mesecno. U ovom izvestajnom periodu na reaktoru RA bilo je zaposleno prosecno 47 radnika, sto je dovoljan broj u uslovima remontnih i investicionih radova. Nedostatak finansijskih sredstava za odrzavanje reaktora RA je neresen problem i u ovom periodu. Ovaj izvestaj podeljen je u dve celine: pogon i odrzavanje reaktora i zastita od zracenja na reaktoru RA.

  12. Research nuclear reactor RA - Annual Report 1990 with the comparative evaluation for the period 1986-1990; Istrazivacki nuklearni reaktor RA - Izvestaj za 1990. godinu uz uporedni pregled za period 1986 - 1990

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1990-12-15

    Annual report concerning the project 'RA research nuclear reactor' for 1990, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1990. godinu, koji finansira republicka zajednica za nauku SR Srbije po ugovoru br. 3705/1 sastoji se iz dva dela. Prvi deo obuhvata pogon i odrzavanje nuklearnog reaktora RA, sto predstavlja obavezu Odeljenja za reaktorski inzenjering u sastavu OOUR Instituta za multidisciplinarna istrazivanja i inzenjering RA. Drugi deo obuhvata poslove zastite od zracenja na reaktoru RA, sto predstavlja obavezu OOUR Instituta za zastitu od zracenja 'Zastita'. Naucno vece Instituta za multidisciplinarna istrazivanja i inzenjjring RA ocenilo je da sadrzina ovog izvestaja odgovara izvrsenim poslovima na reaktoru RA u 1989. godini. Pozitivno se ocenjuje obim i kvalitet izvrsenih radova kako u pogledu odrzavanja i rekonstrukcije reaktora, tako i u pogledu poslova zastite od zracenja izvrsenih kod njega.

  13. RA Research nuclear reactor, Part 1, RA reactor operation and maintenance in 1993, with comparative review for the period 1991 - 1993, Annex 3; Projekat Istrazivacki nuklearni reaktor RA - 1 Deo Pogon i odrzavanje nuklearnog reaktora RA u 1993. godini, uz uporedni pregled za period 1991 - 1993. - prilog 3

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Mikic, N; Tanaskovic, M [Vinca Institute of Nuclear Sciences, Beograd (Serbia and Montenegro)

    1993-12-15

    RA reactor was not operated during 1993 because of the complete instrumentation exchange. Although it has been planned to exchange the complete instrumentation until the end of 1993, and to start reactor operation in the first half of 1993 this was not fulfilled because the instrumentation was not delivered until the end of 1993. Main activities during past seven years were related to construction of the emergency cooling system; repair and reconstruction of the system for handling the spent fuel and improvement of spent fuel storage conditions; exchange of the aged instrumentation. Other reactor components and systems, reactor core, primary coolant loop and gas circulation system are in good condition concerning future start-up. [Serbo-Croat] U 1993. godini reaktor nije bio u pogonu zbog zamene njegove celokupne instrumentacije. Iako je bilo planirano da se celokupna instrumentacija zameni do kraja 1993. te da reaktor pocne sa radom u prvoj polovini 1993. Ovo nije ispunjeno jer celokupna oprema nije isporucena ni do kraja 1993. godine. Osnovni zahvati koji su u proteklih sedma godina izvrseni, odnosili su se na izgradnju sistema za udesno hladjenje, rekonstrukciju sistema za rukovanje ozracenim gorivom i poboljsanje uslova za stokiranje ovog goriva, zamenu instrumentacije. Ostali sistemi reaktora, reaktorsko jezgro, primarno kolo hladjenja i sistem za cirkulaciju gasa su u dobrom stanju i mogu se nesmetano koristiti u buducem radu.

  14. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988; Istrazivacki nuklearni reaktor RA, deo 1, pogon i odrzavanje nukleanog reaktora RA u 1988. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1988-12-15

    According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues. [Serbo-Croat] Prema planu za 1988. godinu, reaktor RA je trebalo da pusten u rad oktobra meseca, medjutim nije dobio dozvolu za nastavak rada. Kontrola i odrzavanje opreme izvrsavani su redovno i efikasno, u granicama koje su diktirane raspolozivoscu repromaterijala i rezervnih delova. Najvecu poteskocu pricinjavalo je odrzavanje instrumentacije. Period stajanja u 1988. godini iskoriscen je za remont teskovodnih pumpi u primarnom kolu hladjenja. U cilju povecanja pouzdanosti rada reaktora zapoceti su radovi na modernizaciji instrumentacije, projekat je izradjen u sovjetskoj organizaciji Atomenergoeksport, sklopljen je ugovor o izradi ove opreme koja bi trebalo da bude isporucena do kraja 1990. U cilju povecanja prostora za skladistenje ozracenog

  15. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor; Istrazivacki nuklearni reaktor RA - Deo 1 - Pogon, odrzavanje i eksploatacija nuklearnog reaktora RA u 1982. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Kozomara-Maic, S; Cupac, S; Radivojevic, J; Stamenkovic, D; Skoric, M; Miokovic, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1982-12-15

    outdated spare parts. Project concerning renewal of the reactor equipment was initiated during the past year according to the contract with the Soviet Atomenergoexport and IAEA which has planned to spend 1 000 000 of rubles for this project. [Serbo-Croat] Probni rad reaktora zapocet septembra 1981. godine na snazi od 2 MW sa 80% obogacenim gorivom nastavljen je u celoj 1982. prema prethodno napravljenom planu. Pocetno jezgro formirano ja sa 44 gorivna kanala sa po deset gorivnih elemenata. Prva polovina godine iskoriscena je za neophodna merenja i ispitivanja radnih parametara rektora i funkcionisanje sistema i opreme u radnim uslovima. U drugoj polovini godine zapocet je program probnog rada na visim snagama. Utvrdjeno je da ugradjeni visak reaktivnosti i kapacitet kontrolnih sipki zadovoljava sigurnosne kriterijume MAAE, ugradjeni visak reaktivnosti moze da omoguci rad na snazi od 4,7 MW u 4 mesecne kampanje sa po 15-20 dana rada, postoje povoljni uslovi za hladjenje jezgra pri pocetnoj konfiguraciji. Izmeren je efekat pocetnog zatrovanja na reaktivnost i raspodelu snage, izmerena je pocetna prostorna raspodela neutronskog fluksa koja iznosi 3,9 10{sup 13} cm{sup -2} s{sup -1} pri znazi od 2 MW. Odredjena je promena kalibracionog koeficijenta u sistemu za automatsko odrzavanje snage. Svi rezultati ukazuju da ce pri nominalnoj snazi od 4,7 Mw biti zadovoljeni svi kriterijum sigurnosti i postovana ogranicenja u odnosu na koriscenje goriva. Po dobijanju dozvole za rad na punoj snazi morace da se izvrsi dopunski probni rad na snagama od 3, 4, i 4,7 MW. Prelaz od pocetne konfiguracije sa 44 gorivna kanala u jezgru vrsice se postupno da bi se dostigla ravnotezna konfiguracija sa 72 gorivna kanala sa po 10 elemenata. Reaktor nije radio u septembu mesecu zbog radova na zameni dela cevovoda koji povezuje pumpnu stanicu na Dunavu sa horizontalnim taloznikom. Kontrola i odrzavanje opreme izvrsavani su redovno u granicama raspolozivosti rezerbinh delova. Teskocu pricinjava

  16. Radiological protection in nucleus reactor; Perlindungan radiologi di reaktor nukleus

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-12-31

    The chapter briefly discussed the following subjects: radiological protection problems of reactor 1. in operation 2. types of reactor i.e. power reactors, research reactors, etc. 3. during maintenance and installation of fuels. 4. nuclear fuels.

  17. The effective lifetime and temperature coefficient in a coupled fast-thermal reactor; Temps de vie effectif et coefficient de temperature dans un reacteur a couplage neutrons rapides-neutrons thermiques; Ehffektivnyj srok zhizni i temperaturnyj koehffitsient nejtronov v dvoyakom reaktore na bystrykh i teplovykh nejtronakh; Vida efectiva y coeficiente de temperatura en un reactor con acoplamiento rapido-termico

    Energy Technology Data Exchange (ETDEWEB)

    Haefele, W. [Kernforschungszentrum, Karlsruhe (Germany)

    1962-03-15

    . Teplovoj komponent dejstvuet kak svoego roda zamedlitel' vremeni zhizni nejtronov. Kak i v teorii zapazdyvayushchikh nejtronov, ehffekt zapazdyvaniya ischezaet, esli reaktivnost' dostatochno vysoka, chtoby bystryj komponent stal kritichnym sam po sebe. V issledovanii rassmatrivalsya sparennyj reaktor, v kotorom bystryj komponent podvergaetsya dejstviyu vnezapnogo stupenchatogo skachka reaktivnosti {alpha}{sub 0}. Izza vozrastayushchego urovnya ehnergii temperatura podnimaetsya i nachinayut rabotat' dva temperaturnykh koehffitsienta: temperaturnyj koehffitsient bystrogo komponenta i temperaturnyj koehffitsient teplovogo komponenta. Ehta problema rassmatrivaetsya s odnoj gruppoj zapazdyvayushchikh nejtronov (v obychnom znachenii). Privoditsya formalizm dlya vyrazheniya ehffektivnogo sroka zhizni i temperaturnogo koehffitsienta vo vremya razlichnykh stadij issledovaniya. Dany takie otkloneniya dlya razlichnykh znachenij {alpha}{sub 0}, pri kotorykh dostigaetsya predel kinetiki reaktorov na bystrykh nejtronakh. (author)

  18. Measurements with a Pulsed and Modulated Source in a Reactor; Mesures au Moyen d'une Source Pulsee et Modulee dans un Reacteur; Izmereniya v reaktore s pomoshch'yu impul'snogo i moduliruemogo is tochnika; Mediciones Efectuadas en Reactor con una Fuente Pulsada y Modulada

    Energy Technology Data Exchange (ETDEWEB)

    Rotter, W. [Centre d' Etude de l' Energie Nucleaire, Mol (Belgium)

    1965-10-15

    analizador multicanal. Como el flujo del generador es perfectamente sinusoidal, la respuesta del reactor puede integrarse en cada cuarto de perfodo, puesto que el circuito de medicion esta gobernado por el generador; por consiguiente, el tiempo de medicion es minimo. Los datos registrados sobre cinta perforada se analizan con ayuda de una calculadora numerica. (author) [Russian] Generator, vyhod nejtronov kotorogo izmenjaetsja v zavisimosti ot funkcii kakogo-to vremeni, byl razrabotan v issledovatel'skih laboratorijah Filipsa. Ego prakticheskaja pol'za v oblasti fiziki reaktorov byla prodemonstrirovana na serii izmerenij, provedennyh v reaktore BR-O 2 v podkriticheskom sostojanii. Horoshaja ustojchivost', vozmozhnost' proizvodit' rezkie izmenenija intensivnosti nejtronnogo potoka, impul'sirovat' vyhod ili modulirovat' ego sinusoidal'no, - vse jeto delaet takoj generator ochen' gibkim. On pozvoljaet ustanavlivat' reaktivnost' ({rho} = {Delta}k/{beta}) i vremja zhizni nejtronov ( Script-Small-L /{beta}) po razlichnym nezavisimym metodam. Tochnoe sravnenie jetih metodov vozmozhno, poskol'ku poslednie mogut byt' ispol'zovany bez izmenenija uslovij izmerenija. Ustanovleno: 1) {rho} na osnove zapazdyvajushhih nejtronov putem mgnovennogo umen'shenija vyhoda nejtronov; 2) {rho} na osnove mgnovennyh nejtronov putem ispol'zovanija impul'sov nejtronov; 3) Script-Small-L /{beta} putem soedinenija 1) i 2) dlja 0,5 $ < {rho} < 2 $ ; 4) Script-Small-L /{beta} na osnove peredatochnoj funkcii reaktora dlja moduliruemogo istochnika. Obsuzhdajutsja peredatochnye funkcii dlja oscilljatora reaktivnosti i dlja sinusoidal'no moduliruemogo istochnika. Pokazano, chto izmerenie Script-Small-L /{beta} vozmozhno dlja 0,1 $< {rho} < $ s uchetom primenenija moduliruemogo istochnika. Tot zhe metod takzhe daet reaktivnost' s pomoshh'ju otnoshenija mgnovennyh nejtronov k zapazdyvajushhim nejtronam dlja optimal'noj chastoty, prichem na praktike jeto proishodit nezavisimo ot dannyh, otnosjashhihsja k

  19. Change of I-V characteristics of SiC diodes upon reactor irradiation; Modification des caracteristiques I-V de jonctions p-n au SiC du fait d'une irradiation dans un reacteur; Izmeneniya kharakteristik I-V vyrashchennogo v SiC perekhoda tipa p-n posle oblucheniya ego v reaktore; Modificaciones que sufren por irradiacion en un reactor las caracteristicas I-V de uniones p-n en SiC

    Energy Technology Data Exchange (ETDEWEB)

    Heerschap, M; De Coninck, R [Solid State Physics Dept., SCK-CEN, Mol (Belgium)

    1962-04-15

    distintas procedencias. Las uniones se obtuvieron en el horno de Lely. Midieron las caracteristicas directa e inversa durante la irradiacion, y despues de esta, hasta la temperatura de 150{sup o}C. Se estan realizando mediciones hasta 500{sup o}C. Han encontrado que uno de los tipos de diodo es resistente a los neutrones del BR-1 hasta un flujo integrado de 10{sup 15} neutrones/cm{sup 2}, mientras que el otro soporta hasta 10{sup 17} neutrones/cm{sup 2}. La memoria indica los cambios de las caracteristicas, asi como los resultados de algunos experimentos de recocido. (author) [Russian] V poiskakh poluprovodnikov, kotorye mogli by byt' ispol'zovany v reaktorakh s vysokoj plotnost'yu nejtronnogo potoka dlya izmereniya raspredeleniya potokov, my obluchali v bel'gijskom reaktore BR-1 perekhody tipa p-n v SiC. Byli oblucheny dva tipa diodov SiC razlichnogo proiskhozhdeniya. EHti perekhody vyrashchivayutsya v pechi Loli. Izmeneniya pryamoj i obratnoj kharakteristik byli izmereny posle i vo vremya oblucheniya vplot' do temperatury 150{sup o}C; v nastoyashchee vremya proizvodyatsya izmereniya vplot' do temperatury 500{sup o}C. Bylo ustanovleno, chto odin tip dioda vyderzhivaet nejtronnoe obluchenie reaktora BR-1 vplot' do integrirovannogo potoka 10{sup 15} nejtronov na kv. sm, togda kak drugoj tip vyderzhivaet obluchenie vplot' do potoka 10{sup 17} nejtronov na kv. sm. Dayutsya izmeneniya kharakteristik, a takzhe rezul'taty nekotorykh ehksperimentov otzhiga. (author)

  20. The Role of Exponential and PCTR Experiments at Hanford in the Design of Large Power Reactors; Roles Respectifs des Experiences Exponentielles et du Reacteur d'Etude des Constantes Physiques de Hanford dans les Etudes de Grands Reacteurs de Puissance; Znachenie ehksponentsial'nykh opytov i opytov na reaktore PCTR pri proektirovanii bol'shikh ehnergeticheskikh reaktorov v khehnforde; Papel de los Experimentos Exponenciales y del Reactor PCTR de Hanford en el Proyecto de Grandes Reactores de Potencia

    Energy Technology Data Exchange (ETDEWEB)

    Heineman, R. E. [General Electric Company, Richland, WA (United States)

    1964-02-15

    introduccion de cambios en el manejo de los reactores ya existentes, como instrumentos de investigacion en la esfera de la fisica de los reactores y como medio de ensenanza. Compara tambien los capitales invertidos en esas instalaciones y los gastos de explotacion. Describe el perfeccionamiento de nuevas tecnicas experimentales que estas instalaciones permiten aplicar con miras a satisfacer la demanda de nuevos datos experimentales. Es menester tener presentes todos estos datos para poder predecir la evolucion de las necesidades y las tendencias futuras en el empleo de estas instalaciones para los estudios de los reactores de potencia. La memoria describe sucintamente el reactor para el estudio de constantes fisicas e indica la manera en que se piensa utilizarlo en el marco de esa evolucion. (author) [Russian] V Hjen- fordskih laboratorijah v techenie pochti 15 let provodjatsja jeksponencial'nye reaktornye iz- merenija na grafito-uranovyh reshetkah. Hotja rezul'taty jetih opytov ispol'zovalis' dlja opredelenija laplasianov predlagaemyh proizvodjashhih reaktorov, oni takzhe sodejstvovali razvitiju ponimanija fiziki reaktorov jetih sistem. Davno priznano, chto poleznost' kri- ticheskogo opyta ogranichena vvidu ego bol'shogo masshtaba i nedostatochnoj chuvstvitel'nosti v otnoshenii nebol'shih lokalizovannyh narushenij sistemy. Zatem mysl' byla napravlena na sozdanie cel'nogo opytnogo reaktora, v kotorom bylo by svedeno do minimuma kolichest- vo materialov, neobhodimyh dlja poluchenija nuzhnyh dannyh. Jeta popytka privela k postrojke usovershenstvovannoj kriticheskoj ustanovki s neskol'kimi zonami reaktora dlja izmerenija fizicheskih konstant PCTR. Ustanovka ispol'zuetsja dlja okazanija sodejstvija pri razrabot- ke proekta po fizike reaktorov dlja neskol'kih jenergeticheskih reaktorov. Krome togo,re- aktor RSTNjavljaetsja ustanovkoj obshhego naznachenija dlja provedenija izmerenij poperechnyh sechenij na reaktore i dlja opredelenija differencial'nyh i integral'nyh fizicheskih para

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

    International Nuclear Information System (INIS)

    Majercik, J.

    1998-01-01

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

  2. UNJUK KERJA REAKTOR PLASMA DIELECTRIC BARRIER DISCHARGE UNTUK PRODUKSI BIODIESEL DARI MINYAK KELAPA SAWIT

    Directory of Open Access Journals (Sweden)

    Ardian Dwi Yudhistira

    2013-10-01

    Full Text Available Biodiesel is one of alternative renewable energy source to substitute diesel fuel. Various biodiesel productionprocesses through transesterification reaction with a variety of catalysts have been developed by previousresearcher. This process still has the disadvantage of a long reaction time, and high energy need. DielectricBarrier Discharge (DBD plasma electro-catalysis may become a solution to overcome the drawbacks in theconventional transesterification process. This process only needs a short time reaction and low energy process.The purpose of this study was to assess the performance of DBD plasma rector in making biodiesel such as: theeffect of high voltage electric value, electrodes gap, mole ratio of methanol / oil, and reaction time. TheResearch method was using GC-MS (Gas Cromatography-Mass Spectrofotometry and FTIR (FourierTransform Infrared Spectrofotometry and then it will be analysed the change of chemical bond betweenreactant and product. So, the reaction mechanism can be predicted. Biodiesel is produced using methanol andpalm oil as reactants and DBD plasma used as reactor in batch system. Then, reactants contacted by highvoltage electric. From the results of this research can be concluded that the reaction mechanism occurs in theprocess is the reaction mechanism of cracking, the higher of electric voltage and the longer of reaction time leadto increasing of product yield. The more of mole ratio of methanol / oil and widening the gap between theelectrodes lead to decreased product yield. From this research, product yield maksimum is 89,8% in the variableof rasio mol metanol/palm oil 3:1, voltage 10 kV, electrode gap 1,5 cm, and reaction time 30 seconds.

  3. Analisis Computational Fluid Dynamics untuk Perancangan Reaktor Gasifikasi Sekam Padi Tipe Downdraft

    Directory of Open Access Journals (Sweden)

    Dziyad Dzulfansya

    2014-10-01

    Full Text Available Rice husk is one of biomass type which can be utilized as gasification’s feedstock for producing combustible gas which can be used as fuel in internal combustion engine. The objective of this research was to obtain the best design of small scale rice husk gasifier from among geometry scenarios by applying computational fluid dynamics method. The geometry scenarios used in this study were angle of throat 70O, 80O, and 90O, and also angle of nozzel 10O and 20O. The softwares used in this study were Gambit 2.4.6 (meshing 3D model and Ansys Fluent 13.0 (simulation. The reactions involved in gasification (3 heterogeneous reactions and 6 homogeneous reactions were solved by finite rate/Eddy dissipation model. Results of simulation showed that gasifier with angle of throat 90O and angle of nozzel 10O produced the highest heating value of gas with volume fraction of CO, H2, and CH4 is 14.49%, 9.65%, and 2.39% respectively. This result showed reasonable agreement with experimental data from other researchers on rice husk gasification.

  4. Research nuclear reactor RA - Annual report 1992; Istrazivacki nuklearni reaktor RA - Izvestaj za 1992. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1992-12-01

    Research reactor RA Annual report for year 1992 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. Second annex B is a paper by Z. Vukadin 'Recurrence formulas for evaluating expansion series of depletion functions' published in 'Kerntechnik' 56, (1991) No.6 (INIS record no. 23024136. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data. [Serbo-Croat] Godisnji izvestaj o radu nuklearnog reaktora RA za 1992. godinu sastoji se od dva dela: prvi deo obuhvata pogon i odzavanje reaktora RA, a drugi poslove zastite od zracenja na reaktoru RA. Prvi deo sadrzi 8 priloga, koji opisuju rad reaktora i poslove sluzbi za odrzavanje opreme i komponenti, finansijski izvestaj, kadrovsku strukturu osoblja reaktora. Drugi prilog (B) je rad Z. Vukadina 'Recurrence formulas for evaluating expansion series of depletion functions' objavljen u casopisu Kerntechnik, 1991. Drugi deo izvestaja o poslovima zastite od zracenja sadrzi 4 priloga sa podacima radijacione kontrole radne sredine i okoline reaktora, opis poslova dekontaminacije i sakupljanja radioaktivnih materija, kao i meteoroloske podatke.

  5. Perhitungan Kebutuhan Cooling Tower Pada Rancang Bangun Untai Uji Sistem Kendali Reaktor Riset

    OpenAIRE

    Awwaluddin, Muhammad; Santosa, Puji; Suwardiyono, Suwardiyono

    2012-01-01

    CALCULATION OF THE NEED FOR COOLING TOWER ON DESIGN OF STRAND TEST RESEARCH REACTOR CONTROL SYSTEM. Cooling tower on the strand test engineering research reactor control system functioning as a heat transfer medium from the heat exchanger to air. To get the transfer of heat or cooling is maximal then the determination of cooling tower needs to be precise. Cooling tower is expected to accept and release heat at 1.191 kw from the heat exchanger. To support these needs will require the calculati...

  6. PENENTUAN KOEFISIEN DISPERSI ATMOSFERIK UNTUK ANALISIS KECELAKAAN REAKTOR PWR DI INDONESIA

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    Full Text Available Atmosfer merupakan pathway penting pada perpindahan radionuklida yang lepas dari Pembangkit Listrik Tenaga Nuklir (PLTN ke lingkungan dan manusia. Penerimaan dosis pada lingkungan dan manusia dipengaruhi oleh sourceterm dan kondisi tapak PLTN. Untuk mengetahui penerimaan dosis lingkungan untuk PLTN di Indonesia, maka diperlukan nilai koefisien dispersi untuk tapak potensial yang dipilih. Model perhitungan dalam penelitian ini menggunakan model yang diterapkan pada paket program pada modul ATMOS dan CONCERN dari PC-Cosyma yaitu model perhitungan segmented plume model. Perhitungan dilakukan untuk PLTN tipe PWR kapasitas 1000 MWe berbahan bakar UO2, postulasi kejadian untuk kecelakaan DBA, kondisi tapak kasar, untuk 6 tapak contoh tapak Semenanjung Muria, Pesisir Banten, dan tapak yang didominasi oleh stabilitas cuaca C,D,E, dan F. Koefisien dispersi dihitung untuk 8 kelompok nuklida produk fisi yang lepas dari PLTN yaitu: kelompok gas mulia, lantanida, logam mulia, halogen, logam alkali, tellurium, cerium, dan kelompok stronsium & barium. Perhitungan input menggunakan paket program ORIGEN-2 dan Arc View untuk penyiapan input perhitungan. Hasil pemetaan untuk parameter dispersi maksimum rerata diperoleh pada jarak radius 800 m dari sumber lepasan untuk nuklida dari kelompok logam mulia, logam alkali dan kelompok nuklida cerium. Parameter dispersi untuk Tapak Muria maksimum 1,53E-04 s/m3, Tapak Serang adalah 1,40E-03 s/m3, tapak dengan stabilitas C: 1,72E-04 s/m3, stabilitas D: 1,40E-04 s/m3, Stabilitas E: 1,07E-04 s/m3, dan tapak dengan stabilitas F : 2,14E-05 s/m3. Kata kunci: koefisien dispersi, atmosferik, PWR, kecelakaan, Indonesia   The atmosphere is an important pathway in the migration of radionuclides transport from the Nuclear Power Plant (NPP to the environment and humans. The dose accepted in the environment and humans is influenced by the sourceterm and NPP siting condition. Distribution of radionuclides in the atmosphere is determined by the dispersion coefficient. To find the environment dose acceptance for nuclear power plants in Indonesia, it is necessary to map the dispersion coefficient for Indonesia potential siting Model calculations in this study using Segmented plume model, which a model that is applied to the ATMOS and CONCERN module of PC-Cosyma software. The calculation has done for PWR 1000 MWe with UO2 fuel, DBA accident postulations, roughnes site conditions, for 8 example site such as Muria Peninsula, Coastal Banten, and the C, D, E, and F stability. Dispersion coefficient was calculated for the 8 fission product groups are: the noble gases, lanthanides, noble metals, halogens, alkali metals, tellurium, cerium, and strontium & barium groups. Input calculation using the program package Origen-2 and Arc View for the preparation of input calculations. The results of the dispersion parameter calculated are: the average maximum is obtained at a distance of 800 m radius from the source, for noble metals, alkali metal and cerium group nuclides. Dispersion parameters for maximum at Muria site is 1.53E-04 s/m3, Serang site is 1.40E-03 s/m3, site with stability C is 1.72E-04 s/m3, stability D is 1.40E-04 s/m3, stability E is 1.07E-04 s/m3, and site with the stability F is 2.14E-05 s/m3. Keywords: dispersion coefficient, atmospheric, PWR, accident, Indonesia

  7. Start-up and Performance Characteristics of a Trickle Bed Reaktor Degrading Toluene

    Czech Academy of Sciences Publication Activity Database

    Misiaczek, O.; Paca, J.; Halecký, M.; Gerrard, A. M.; Sobotka, Miroslav; Soccol, C. R.

    2007-01-01

    Roč. 50, č. 5 (2007), s. 871-877 ISSN 1516-8913 Grant - others:GA ČR(CZ) GA104/05/0194 Institutional research plan: CEZ:AV0Z50200510 Source of funding: V - iné verejné zdroje Keywords : biotrickling filter * constructed mixed population * toluene Subject RIV: EE - Microbiology, Virology Impact factor: 0.349, year: 2007

  8. RA research reactor - properties and experimental capabilities; Istrazivacki reaktor RA - Tehnicke karakteristike i eksploatacione mogucnosti

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M; Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1978-05-15

    The brief survey of the Reactor RA exploitation experience, as well as the reactor equipment state, after 18 years of operation is presented. The results of efforts spent on reactor characteristics improvement in order to ensure safe and reliable reactor operation for next 15-20 years, are described. Prikazani su fragmenti iz eksploatacije reaktora kao i stanje opreme, posle 18 godina rada. Na kraju je dat prikaz sta je preduzeto i sta se preduzima da se poboljsaju karakteristike i poveca sigurnost i bezbednost rada za sledecih 15-20 godina.

  9. Reproduction of the RA reactor fuel element fabrication; Reprodukcija izrade gorivnog elementa za reaktor RA

    Energy Technology Data Exchange (ETDEWEB)

    Novakovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This document includes the following nine reports: Final report on task 08/12 - testing the Ra reactor fuel element; design concept for fabrication of RA reactor fuel element; investigation of the microstructure of the Ra reactor fuel element; Final report on task 08/13 producing binary alloys with Al, Mo, Zr, Nb and B additions; fabrication of U-Al alloy; final report on tasks 08/14 and 08/16; final report on task 08/32 diffusion bond between the fuel and the cladding of the Ra reactor fuel element; Final report on task 08/33, fabrication of the RA reactor fuel element cladding; and final report on task 08/36, diffusion of solid state metals. Ovaj rad sadrzi devet priloga: 1. Zavrsni izvestaj o podzadatku 08/12, ispitivanje elementa goriva reaktora RA; 2. Koncepcija izrade gorivnog elementa reaktora RA; 3. Ispitivanje mikrostrukture gorivnog elementa reaktora RA; 4. Zavrsni izvestaj o podzadatku 08/13, dobijanje binarnih legura urana sa legirajucim komponentama Al, Mo, Zr, Nb i B; 5. Dobijanje legure U-Al; 6. Zavrsni izvestaj o podzadacima 08/14 i 08/16; 7. Zavrsni izvestaj o podzadatku 08/32, difuziona veza goriva i kosuljice gorivnog elementa reaktora RA; 8. Zavrsni izvestaj o podzadatku 08/33, izrada kosuljice gorivnog elementa reaktora RA; 9. Zavrsni izvestaj o podzadatku 08/36, difuzija kod metala u cvrstom stanju.

  10. PENGOLAHAN LIMBAH CAIR ORGANIK SECARA BIOLOGI MENGGUNAKAN REAKTOR ANAEROBIK LEKAT DIAM

    OpenAIRE

    Indriyati, Indriyati

    2018-01-01

       Organic waste water can be treated biolocally by using anaerobic fixed bed reactor. Fixed bed reactor is bioreactor which is compleeted with support material inside reactor for bacteria fixation in the surface area of support material. The benefit of using this kind of technology are it needs low energy, low nutrien, low sludge production and could treat high organic concentraion waste water.   The support material  has important role in the  Fixed Bed reactor performance, therefore it mus...

  11. Research reactor 'A' 6.5/10 MW; Istrazivacki reaktor 'A' 6,5/10 MW

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1961-07-01

    This booklet includes a short description of the RA research reactor with basic properties of its components: control and safety system, heavy water system, technical water cooling system, heavy water distillation system, cover gas system, dosimetry control system, power supply system. It is used for fundamental research in reactor and nuclear physics, isotope production, materials testing.

  12. RB Research nuclear reactor, 30 years of operation; Istrazivacki nuclearni reaktor RB, povodom 30 godina rada

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M; Stefanovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1988-06-15

    Paper describes utilization, modifications and changes of construction and control-safety systems done at the RB reactor during 30 years of operation. Experiments performed at the reactor are summarized, new reactor equipment is described and the future plans are shown. Rad prikazuje eksploataciju reaktora RB tokom 30 godina rada, modifikacije i izmene u konstruktivnim i upravljacko-sigurnosnim sistemima. Sumirani su eksperimenti izvedeni na njemu, prikazana je nova oprema i planovi za buduci rad.

  13. Nuclear RB research reactor. Thirty years of anniversary; Istrazhivacki nuklearni reaktor RB. Povodom 30 godina rada

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M; Stefanovic, D [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia)

    1988-07-01

    Nuclear research reactor RB in the Nuclear Engineering Laboratory - NET at the 'Boris Kidric' Institute of Nuclear Sciences in Vinca is the first reactor system built in Yugoslavia in 1958. This year is the thirtieth anniversary of the RB reactor operation, which has survived a series of modifications trying to follow a contemporary nuclear research directions. This report describes its basic technical characteristics and experimental possibilities. Especially, the modifications in the last 25 years are underlined, the experiences gained, and new plans for the future are presented. (author)

  14. Investigations of the chemical states of carrier-free phosphorus-32 as extracted into water from pile-irradiated sulphur; Recherches sur les etats chimiques du phosphore-32 sans entraineur obtenu par extraction aqueuse a partir de soufre irradie dans un reacteur; Issledovanie khimicheskogo sostoyaniya svobodnogo ot nositelya fosfora-32 pri izvlechenii ego v vodu iz obluchennoj v yadernom reaktore sery; Estudio de los estados quimicos del fosforo-32 libre de portador que se obtiene por extraccion acuosa del azufre irradiado en un reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dahl, J B; Birkelund, O R [Institutt for Atomenergi, Kjeller, Lillestrom (Norway)

    1962-01-15

    {>=} 5). Aucun metaphosphate (cyclique) n'a ete decele dans la solution lors de la production et du stockage. Les resultats indiquent que les composes de polyphosphore ont ete formes dans le materiau de cible au cours de l'irradiation. On a etudie plus particulierement l'adsorption des composes du phosphore-32 sans entraineur par le verre dans les conditions de l'experience. (author) [Spanish] Uno de los metodos que se emplean para obtener fosforo-32 libre de portador consiste en irradiar azufre en un reactor y extraer con agua el fosforo formado. La presente memoria informa sobre los estados quimicos del fosforo-32 en soluciones acuosas durante las diversas fases del proceso normal de obtencion. Los autores han estudiado tambien las modificaciones que el estado quimico de los compuestos de fosforo-32 experimenta en el producto final, en funcion del tiempo de almacenamiento. Han encontrado que el fosforo-32 aparece principalmente en forma de orto-fosfato. La proporcion de ortofosfato aumenta durante el tratamiento quimico; al comenzar la extraccion, es del orden del 70 por ciento, en tanto que alrededor del 98 por ciento del fosforo-32 libre de portador obtenido como producto final se encuentra bajo forma de ortofosfato. Los componentes restantes de la mezcla consisten en piro-, tri-, tetra- y polifosfatos de cadena larga (con un numero de atomos de fosforo {>=} 5). Durante la elaboracion y almacenamiento los autores no han encontrado metafosfatos (ciclicos) en ninguna de las soluciones. Los resultados obtenidos indican que los compuestos polifosforicos se formaron en el material del blanco durante la irradiacion. Los autores han prestado especial atencion al estudio de la adsorcion de compuestos de fosforo-32 libre de portador por el material de vidrio en las condiciones experimentales. (author) [Russian] Odnim iz sovremennykh metodov polucheniya svobodnogo ot nositelya fosfora-32 yavlyaetsya izvlechenie ego v vodu iz obluchennoj v yader- nom reaktore sery. V

  15. RB research nuclear reactor, Annual report for 1981; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1981. godini

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Sotic, O; Pesic, M; Vranic, S; Zivkovic, B; Bogdanovic, M; Petronijevic, M [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1981-07-01

    The annual report for 1981 includes the following: utilization of the RB reactor; accident and incidents analysis; description of the reactor equipment status; dosimetry and radiation protection; RB reactor staff; financial data. Seven Annexes to this report are concerned with: maintenance of the reactor components and equipment, including nuclear fuel, heavy water, reactor vessel, heavy water coolant circuit, experimental platforms, absorption rods; maintenance of the electric power supply system, neutron source equipment, crane; control and maintenance of ventilation and heating systems, gas and comprised gas systems, fire protection system; plan for renewal of the reactor components; contents of the RB reactor safety report; reactor staff; review of measured radiation doses; experimental methods; training of the staff; and financial report.

  16. RB research nuclear reactor, Annual report for 1982; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1981. godini

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Zivkovic, B [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1982-12-15

    This report includes data concerned with reactor operation and utilization, status of reactor components and equipment, refurbishment of the equipment, dosimetry and radiation protection, reactor staff, financing. It includes 7 Annexes as follows: Maintenance of reactor equipment in 1982; contents of the RB reactor safety report; review of radiation doses in the reactor building and exposure doses for the reactor staff; utilization of the RB reactor in 1982; and financial data.

  17. Passive device for emergency core cooling of pressurized water reactors. Pasivno ustrojstvo za bezopasnost na vodo-voden atomen reaktor

    Energy Technology Data Exchange (ETDEWEB)

    Sikora, D

    1984-02-28

    The device proposed ensures additional margin of reactor subcriticality in case of post-accident emergency core cooling (ECC), using concentrated solution of chemical absorber and hot water from the secondary circuit. It consists of: a) a differential cylinder with a differential piston in it, with a lid and a seal, connected to a pipeline for secondary coolant; b) a pipeline for the secondary coolant; c) a volume between the lid and the piston for the secondary coolant from the steam generator; d) a discharge pipeline with a check valve of seal type connecting the inner volume of the differential cylinder to the discharge line; and e) a pipeline from the high-pressure volume of the differential cylinder filled with concentrated chemical absorber solution, to one of the main circulation loops. The device permits ECC innovation of the operating non-standard nuclear power plants with PWR type reactors.

  18. Nový úkol hybridních reaktorů – likvidace jaderného odpadu

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2009-01-01

    Roč. 57, č. 11 (2009), s. 24-a33 ISSN 0040-1064 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * ITER * hybrid * nuclear waste * LIFE * FFTS * CFNS * SXD * FDF Subject RIV: BL - Plasma and Gas Discharge Physics

  19. Possibilities of using metal uranium fuel in heavy water reactors; Mogucnosti upotrebe metalnog urana kao goriva za teskovodne reaktore

    Energy Technology Data Exchange (ETDEWEB)

    Djuric, B; Mihajlovic, A; Drobnjak, Dj [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions. Postoje ozbiljni ekonomski razlozi za primenu metalnog urana u teskovodnim reaktorima, pre svega zbog njegove velike gustine, odnosno visokog konverzionog faktora, i zbog niskih troskova proizvodnje gorivnih elemenata. Glavne prepreke su bubrenje pri velikim stepenima sagorevanja i opasnost od korozije. Postoje veliki izgledi da se primenom odredjenih projektnih koncepcija i upotrebom legiranjem poboljsanog urana postigne zadovoljavajuca stabilnost metalnog urana u uslovima rada reaktora (author)

  20. German safety engineering in comparison to the Chernobyl reactor; Deutsche Sicherheitstechnik im Vergleich zum Tschernobyl-Reaktor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-04-15

    In Germany the Russian RBMK would not have been licensed, since the safety standards in Germany were superior to those in the former USSR. The most significant differences - the containment, the control rod system, the void coefficient and the emergency cooling system are shortly summarized. The consequences for the population are cumulative environmental radiation exposure are reported using the official data of IAEA, WHO and UNDP.

  1. Research Project 'RB research nuclear reactor' (operation and maintenance), Final report; Naucnoistrazivacki projekt 'Istrazivacki nuclearni reaktor RB, (pogon i odrzavanje), Zavrsni elaborat projekta

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    This final report covers operation and maintenance activities at the RB reactor during period from 1981-1985. First part covers the RB reactor operation, detailed description of reactor components, fuel, heavy water, reactor vessel, cooling system, equipment and instrumentation, auxiliary systems. It contains data concerned with dosimetry and radiation protection, reactor staff, and financial data. Second part deals maintenance, regular control and testing of reactor equipment and instrumentation. Third part is devoted to basic experimental options and utilization of the RB reactor including training.

  2. RB research nuclear reactor, Annual report for 1984, I - III; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1984. godini, I - III

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Zivkovic, B; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1984-07-01

    The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source.

  3. RB research nuclear reactor - Annual report for 1986, I - III; Istrazivacki nuklearni reaktor RB (Izvestaj o radu u 1986. godini), I-III

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Jevremovic, M; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1987-07-01

    This report includes data concerning the RB reactor operation in 1986, state of the reactor components, data about the employed personnel and the database of experimental and other reactor related devices. It is made of 3 parts: Engineering description and operation of the RB reactor including dosimetry, reactor staff data and financial report; Reactor facility components and maintenance; RB reactor operation and utilization in 1986. Izvestaj pokazuje podatke o radu reaktora RB u toku 1986. godine, stanje reaktorske opreme, podatke o angazovanom osoblju na reaktoru i datoteku sa podacima o eksperimentalnoj i drugoj opremi reaktora RB. Sastoji se od 3 dela: tehnicki opis, pogon i rad reaktora, oprema postrojenja i njeno odrzavanje, koriscenje reaktora u 1986. godini.

  4. Základy fúzní energetiky II - Základní fyzika fúzních reaktorů

    Czech Academy of Sciences Publication Activity Database

    Entler, Slavomír; Mlynář, Jan; Dostál, V.

    Srpen (2016), č. článku 14538. ISSN 1801-4399 Institutional support: RVO:61389021 Keywords : Nuclear fusion * fusion reactors * plasma Subject RIV: JF - Nuclear Energetics http://energetika.tzb-info.cz/elektroenergetika/14538-zaklady-fuzni-energetiky-ii-zakladni-fyzika-fuznich-reaktoru

  5. RB Research nuclear reactor, Annual report for 1994, I - III; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1994. godini, I - III

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovic, D; Milosevic, M; Pesic, M [Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia); Marinkovic, P [Elektrotehnicki fakultet, beograd (Yugoslavia); Kocic, A; Ilic, R; Dasic, N; Ljubenov, V; Petronijevic, M; Jevremovic, M [Institute of Nuclear Sciences Vinca, Belgrade (Serbia)

    1994-12-15

    Report on RB reactor operation during 1994 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization as well as operation of the VAX-8250 computer.

  6. RB research nuclear reactor, Annual report for 1989, I - III; Istrazivacki nukleani reaktor RB (Izvestaj o radu u 1989. godini), I - III

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovic, D; Pesic, M; Hadimahmutovic, N; Vranic, S; Petronijevic, M; Jevremovic, M; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1989-12-15

    This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989.

  7. RB Research nuclear reactor, Annual report for 1995, I-IV; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1995. godini, I-IV

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovic, D; Milosevic, M; Pesic, M [Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia); Marinkovic, P [Elektrotehnicki fakultet, Beograd (Yugoslavia); Ilic, R; Dasic, N; Milovanovic, S; Ljubenov, V; Petronijevic, M; Jevremovic, M [Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1995-12-15

    Report on RB reactor operation during 1995 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor.

  8. Control of a reactor for conditioning of biogenic waste materials. Final report; Steuerung eines Reaktors zur Aufbereitung von Abfaellen mit biogenen Bestandteilen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Bartha, B.; Brummack, J.; Kloeden, W.

    2002-10-01

    Mechanical-biological waste treatment implies biological drying and mechanical separation; it produces a pollutant-free fuel fraction with good calorific value and inorganic fractions available for recycling. Apart from some prototypes, mostly static aerobic reactors are used. For moist residues, dynamic reactors may be interesting as they permit coupling of thermal and mechanical processes, i.e. parallel biological drying and mechanical separation. A discontinuous rotary kiln reactor for biological drying of residues was developed in this project. (orig.) [German] Die mechanisch-biologische Aufbereitung von Restabfaellen (MBA) hat sich zu einem Systembaustein innerhalb der Restabfallentsorgung entwickelt. Das Hauptanwendungsziel besteht darin, durch die Kombination von biologischer Trocknung mit mechanischen Stofftrennverfahren eine heizwertreiche und schadstoffentfrachtete Fraktion mit Brennstoffeigenschaften, sowie verwertbare anorganische Fraktionen aus dem Restabfall zu gewinnen. Die Ausgangssituation zeigte, dass, abgesehen von einigen prototypischen Anlagen, in der MBA-Technologie hauptsaechlich statische, aerob arbeitende Reaktoren (Rottetunnel und Rotteboxen) fuer den biologischen Schritt eingesetzt werden. Fuer feuchte Restabfaelle erscheint es interessant, dynamische Reaktoren einzusetzen, die die Kopplung thermischer und mechanischer Prozesse erlauben, die also die biologische Trocknung und den mechanischen Aufschluss parallel zulassen. Im Rahmen des Projektes wurde die Steuerung eines diskontinuierlich betriebenen Drehrohrreaktors zur biologischen Trocknung von Restabfaellen entwickelt. (orig.)

  9. Research nuclear reactor RA - Annual Report 1975. Operation and maintenance; Istrazivacki nuklearni reaktor RA - Izvestaj za 1975. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-01-15

    The plan for 1975 was successfully fulfilled. This is reflected in research related to improvement of operating properties of the RA reactor, mostly due to the effort of the RA staff employed in operation and maintenance of the reactor. Fuel saving achieved by this activity amounted to about 38% (80% enriched fuel). Preliminary work is done, concerned with new reactor core with highly enriched fuel. This is a significant saving as well. New fuel elements have arrived at the end of this year. It is going to enable increase of neutron flux by 50% without changing the nominal operating power. The possibility of further improvement of the reactor are analyzed, to enable material testing and production of radioactive sources. Mid term plan for reactor operation was made according to this analysis. It is planned to further increase the neutron flux in isolated smaller zones, and building new experimental loops with cooling and fast neutron converters. Much was done to increase the safety level of reactor operation and preparing the safety report. [Serbo-Croat] Izvrsenje zadataka u 1975. godini bilo je uspesno. To se ogleda u povecanju istrazivackog rada vezanog za poboljsanje eksploatacionih karakteristika reaktora RA, pretezno koriscenjem sopstvenog kadra angazovanog u pogonu i odrzavanju reaktora. Ovim radom postignuta je usteda goriva od oko 38% (80% obogaceno gorivo). Izvrseni su preliminarni radovi na prevodjenju reaktora RA na novo gorivo, sto je takodje velika usteda. Novo gorivo je stiglo krajem godine i ono ce obezbediti porast neutronskog fluksa od 50%, bez promene nominalne snage reaktora. Izvrsena je analiza mogucnosti daljeg usavrsavanja reaktora za potrebe ispitivanja materijala kao i proizvodnju radioaktivnih izvora. Na osnovu ove analize nacinjen je srednjorocni program rada reaktora RA sa tezistem na daljem povecanju fluksa u izdvojenim manjim zonama i ugradnju 'hladjenih petlji' i brzih konvertora. Mnogo je ucinjeno na povecanju stepena sigurnosti reaktora RA i izradi sigurnosnog izvestaja.

  10. The Siemens-Argonaut reactor as a driver zone for a high-temperature reactor cell. Der Siemens-Argonaut-Reaktor als Treiberzone fuer eine Hochtemperaturreaktorzelle

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H; Schuerrer, F; Ninaus, W; Oswald, K; Rabitsch, H; Kreiner, H [Technische Univ., Graz (Austria). Inst. fuer Theoretische Physik und Reaktorphysik; Neef, R D [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung

    1984-12-15

    To enable a validation of neutron physics calculation methods for pebble bed reactors the inner reflector of an Argonaut research reactor was substituted by a full of about 1200 fuel elements of the AVR-Juelich type. The report describes the measuring instruments and the reactor physical layout of the arrangement by the code packages GAMTEREX, ZUT-D.G.L. and MUPO. Comparison of calculated reaction rates with measurements show good agreement. Application of the codes to high-temperature reactors in abnormal states is envisaged. (Author, translated by G.Q.)

  11. Studi Model Benchmark Mcnp6 Dalam Perhitungan Reaktivitas Batang Kendali Htr-10

    OpenAIRE

    Jupiter S.Pane, Zuhair, Suwoto, Putranto Ilham Yazid

    2016-01-01

    STUDI MODEL BENCHMARK MCNP6 DALAM PERHITUNGAN REAKTIVITAS BATANG KENDALI HTR-10. Dalam operasi reaktor nuklir, sistem batang kendali memainkan peranan yang sangat penting karena didesain untuk mengendalikan reaktivitas teras dan memadamkan reaktor. Nilai reaktivitas batang kendali harus diprediksi secara akurat melalui eksperimen dan perhitungan. Makalah ini mendiskusikan model Benchmark dalam perhitungan reaktivitas batang kendali reaktor HTR-10. Perhitungan dikerjakan dengan program transpo...

  12. New renewable energy sources and efficient power consumption at NOK: Promoting - testing - informing. Neue erneuerbare Energien und rationelle Stromanwendung bei den NOK: Foerdern - Pruefen - Informieren

    Energy Technology Data Exchange (ETDEWEB)

    Kueffer, K; Baumberger, H [Nordostschweizerische Kraftwerke AG (NOK), Baden (Switzerland)

    1991-09-18

    In accordance with the operating principles, the Nordostschweizerische Kraftwerke (NOK) has decided on a comprehensive programme dedicated to opening up new sources of renewable energy and also to reducing demand by means of efficient power consumption. Intensive promotion, testing and information should ensure that, on the one hand, a premature condemnation of both possibilities and, on the other hand, exaggerated hopes can be avoided. (orig.).

  13. RA Research nuclear reactor, Part II - radiation protection at the RA nuclear reactor in 1984; Istrazivacki nuklearni reaktor RA - Deo II - zastita od zracenja kod nuklearnog reaktora RA u 1984. godini

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M; Ajdacic, N; Zaric, M; Vukovic, Z [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1984-12-15

    Radon protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Collecting and treatment of fluid effluents; and (4) radioactive wastes, decontamination and actions. Each category is described as a separate annex of this report.

  14. Nuclear power in space. Use of reactors and radioactive substances as power sources in satellites and space probes; Kaernkraft i rymden. Anvaendningen av reaktorer och radioaktiva aemnen som kraftkaellor i satelliter och rymdsonder

    Energy Technology Data Exchange (ETDEWEB)

    Hoestbaeck, Lars

    2008-11-15

    Today solar panels are the most common technique to supply power to satellites. Solar panels will work as long as the power demand of the satellite is limited and the satellite can be equipped with enough panels, and kept in an orbit that allows enough sunlight to hit the panels. There are various types of space missions that do not fulfil these criteria. With nuclear power these types of missions can be powered regardless of the sunlight and as early as 1961 the first satellite with a nuclear power source was placed in orbit. Out of seventy known space missions that has made use of nuclear power, ten have had some kind of failure. In no case has the failure been associated with the nuclear technology used. This report discusses to what degree satellites with nuclear power are a source for potential radioactive contamination of Swedish territory. It is not a discussion for or against nuclear power in space. Neither is it an assessment of consequences if radioactive material from a satellite would reach the earth's surface. Historically two different kinds of Nuclear Power Sources (NPS) have been used to generate electric power in space. The first is the reactor where the energy is derived from nuclear fission of 235U and the second is the Radioisotope Thermoelectric Generator (RTG) where electricity is generated from the heat of naturally decaying radionuclides. NPS has historically only been used in space by United States and the Soviet Union (and in one failing operation Russia). Nuclear Power Sources have been used in three types of space objects: satellites, space probes and moon/Mars vehicles. USA has launched one experimental reactor into orbit, all other use of NPS by the USA has been RTG:s. The Soviet Union, in contrast, only launched a few RTG:s but nearly forty reactors. The Soviet use of NPS is less transparent than the use in USA and some data published on Soviet systems are more or less well substantiated assessments. It is likely that also future space probes, moon and Mars vehicles will be using NPS. Besides the more established users of NPS in space, USA and Soviet Union (today Russia), it is possible that we in a not to distant future will see use of NPS in space by China, India and maybe also ESA (European Space Agency). In 1992 the United Nations General Assembly adopted a resolution regarding principles for the use of NPS in space. The resolution consists of eleven points regarding definitions and usage of NPS in space, and how to handle notification and compensation in case of damages due to a failure involving a satellite with an NPS. The probability of radioactive fallout in Sweden following an incident with a NPS-equipped satellite is very small. Due to the fact that everything placed in orbit around Earth sooner or later will re-enter, it is not possible to use probability of re-entry at any time as a measure of risk. Instead the measure Probability of re-entry within 100 year has been chosen. If the routine use of NPS in Low Earth Orbit (LEO) is not taken up again two cases can be defined: - Within about 3 000 years all satellites stored in Nuclear Safe Orbit (NSO) will de-orbit and re-enter the Earth atmosphere. One satellite, Triad OI-IX is in orbit at a lower altitude, and will thus de-orbit earlier. The probability that it does re-enter within 100 years from now is so small that a quantitative measure is deemed not to be meaningful - There is a risk of a launch failure involving a satellite or space probe with a NPS, with a risk of fallout in Sweden. This is not a large risk, but it is orders of magnitude higher than the probability of a satellite that now is in NSO will end up in Sweden within 100 years. If the routine use of NPS in LEO is re-established, the probabilities above are no longer valid

  15. Announcement of recommendations of the Reaktor-Sicherheitskommission. As of 24 July 1997. Joint recommendations of RSK and GPR for safety requirements of future nuclear PWR-type power plants. English versions published in the years 1995 through 1997

    International Nuclear Information System (INIS)

    1997-01-01

    The recommendations, most parts given in English, refer to the European Pressurized Water Reactor (EPR) and have been established by the German RSK (reactor safety commission), the corresponding French organization GPR and the German SSK (radiation protection commission). This publication continues earlier joint recommendations by the national bodies, last published by the German BMU (responsible German ministry) on 5 May 1995, in BAnz. page 7452. The safety recommendations establish the basis for further activities in the Franco-German project for development of the EPR, a PWR type reactor of the next generation. (CB) [de

  16. Project RA Research nuclear reactor - Annual report 1993 with comparative review for the period 1991 - 1993; Projekat Istrazivacki nuklearni reaktor RA - Izvestaj za 1993. godinu, uz uporedni pregled za period 1991 - 1993

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-12-15

    Research reactor RA Annual report for year 1993 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. The ninth separate annex deals with the feasibility of RA reactor applications. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data. [Serbo-Croat] Godisnji izvestaj o radu nuklearnog reaktora RA za 1993. godinu sastoji se od dva dela: prvi deo obuhvata pogon i odzavanje reaktora RA, a drugi poslove zastite od zracenja na reaktoru RA. Prvi deo sadrzi 8 priloga, koji opisuju rad reaktora i poslove sluzbi za odrzavanje opreme i komponenti, finansijski izvestaj, kadrovsku strukturu osoblja reaktora. Poseban prilog razmatra mogucnosti koriscenja istrazivackog reaktora RA. Drugi deo izvestaja o poslovima zastite od zracenja sadrzi 4 priloga sa podacima radijacione kontrole radne sredine i okoline reaktora, opis poslova dekontaminacije i sakupljanja radioaktivnih materija, kao i meteoroloske podatke.

  17. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986; Istrazivacki nuklearni reaktor RA, deo 1, pogon i odrzavanje nukleanog reaktora RA u 1986. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1986-12-01

    In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues. [Serbo-Croat] Sa ciljem da se obezbedi pouzdan rad reaktora RA a u skladu sa zakonskim propisima, zavrsena su tri velika zahvata zapoceta 1984: izgradnja novog sistema za udesno hladjenje, rekonstrukcija postojeceg sistema za ventilaciju, i modernizacija reaktorske instrumentacije. Istovremeno tokom 1985/1986. zapoceta je modernizacija instrumentacije i rekonstrukcija sistema za rukovanje i skladistenje iskoriscenog goriva u zgradi reaktora. Projekti za navedene radove su vec zavrseni ili su u zavrsnoj fazi, a ocekuje se da ce rekonstrukcija oba sistema biti zavrsena do kraja 1988. odnosno sredine 1989. godine. Izrada izvestaja o sigurnosti reaktora RA, prema preporukama MAAE zavrsena je 1986. Investiciona ulaganja na reaktoru Ra u 1986. iskoriscena su za: nabavku 8000 kg teske vode, za investiciono odrzavanje reaktorskih sistema i nabavku opreme, za rekonstrukciju reaktorskih sistema. Ovaj izvestaj sadrzi 8 priloga koji opisuju rad reaktora, rad strucnih sluzbi i finansiranje.

  18. The development of fast reactors - Effects on the Swedish system of management of spent fuel; Utveckling av snabba reaktorer - Paaverkan paa det svenska systemet foer hantering av anvaent braensle

    Energy Technology Data Exchange (ETDEWEB)

    Hans Forsstroem, Hans [SKB International AB, Stockholm (Sweden)

    2013-09-15

    Since the start of the nuclear power era studies have been performed of how to utilise the uranium energy resource in the most effective way. Only about one percent of the energy potential of uranium is utilised in the light water reactors of today. To improve the utilization other types of reactors are needed. With fast reactors theoretically 50-100 times more energy can be extracted from the uranium. This will require reprocessing of the uranium and multiple recycling of the plutonium. Plutonium and uranium can also be recycled in light water reactors, but this will only improve the uranium utilisation by about 20 %. Recycling of plutonium on a routine basis is presently only done in France. The development of fast reactors has been going on since the end of the 1940ies. During the 1970ies the planning was that a large number of fast reactors and their associated fuel cycle facilities would be in operation by the year 2000. The development has, however, for different reasons been much slower than planned. The general assessment today is that fast reactor, if they will be realised, will hardly give an important contribution to energy production until after 2050. Nuclear power production has instead been dominated by light water reactors similar to the ones in use in Sweden. Light water reactors are believed to continue to dominate during the next decades. To start a fast reactor system plutonium (or highly enriched uranium) will be needed. Such plutonium is contained in spent nuclear fuel from light water reactors. This raises the question: Should the spent nuclear fuel be stored so that the potential energy resource in the fuel can be used in the future instead of disposing of it as a waste? The answer to this question will depend on when the material will be useful, i.e. when fast reactors have been introduced on a large scale. It will also depend on the demand for plutonium at this time, i.e. will plutonium be a scarce redundant resource at this point of time. In this context it should be considered that fast reactors will generate their own plutonium, as breeder reactors. Plutonium from other reactors will thus only be needed for the first years of operation. To provide a basis for the answer to the question if the Swedish spent fuel is a resource or a waste this report provides an overview of the present development status for fast reactors and their potential for large scale commercial use. It further describes the impact on the Swedish system for management of spent nuclear fuel if the fuel were to be reprocessed and the uranium and plutonium reused as fuel for fast reactors or for the present reactors.

  19. PEMBUATAN BAHAN BAKU SPREADS KAYA KAROTEN DARI MINYAK SAWIT MERAH MELALUI INTERESTERIFIKASI ENZIMATIK MENGGUNAKAN REAKTOR BATCH [Preparation of Red Palm Oil Based-Spreads Stock Rich in Carotene Through Enzymatic Interesterification in Batch-type Reactor

    Directory of Open Access Journals (Sweden)

    Nur Wulandari1,2

    2012-12-01

    Full Text Available Enzymatic interesterification of red palm oil (a mixture of red palm olein/RPO and red palm stearin/RPS in 1:1 weight ratio and coconut oil (CNO blends of varying proportions using a non-specific immobilized Candida antartica lipase (Novozyme 435 was studied for the preparation of spread stock. The interesterification reaction was held in a batch-type reactor. Two substrate blends were chosen for the production of spread stock i.e. 77.5:22,5 and 82.5:17.5 (RPO/RPS:CNO, by weight through enzymatic interesterification in three different reaction times (2, 4, and 6 hours. The interesterification reactions were conducted at 60°C, 200 rpm agitation speed and 10% of Novozyme 435. The interesterified products were evaluated for their physical characteristics (slip melting point or SMP and solid fat content or SFC and chemical characteristics (carotene retention, moisture content, and free fatty acid/FFA content. All of the interesterified products had lower SFC and SMP as compared to the initial blends. The SMP and SFC increased in longer reaction times. The SMP ranged from 30.8°C to 34.9°C. The carotene retention ranged from 74.80% to 81.08%, while the moisture content and FFA content increased in longer reaction times. The interesterified products had desirable physical properties for possible use as a spread stock rich in carotene.

  20. Slug-Burst Detection in the G3 Reactor; La detection de rupture de gaine au reacteur G3; Obnaruzhenie razryva obolochki v reaktore G3; Deteccion de fallas del revestimiento en el reactor G3

    Energy Technology Data Exchange (ETDEWEB)

    Plisson, J. [Centre d' Etudes Nucleaires de Marcoule (France)

    1963-10-15

    The author explains the principles underlying slug-burst detection and describes the construction of the apparatus concerned. The main features are a) fully automatic operation, b) centralization of data in the control room and c) measurement by electrostatic collection on a turntable. (author) [French] Dans ce memoire, l'auteur expose les principes sur lesquels est fondee la detection de rupture de gaines et il decrit la realisation des installations. Les caracteristiques principales sont a) l'automatisme integral, b) la centralisation des informations dans la salle de commande et c) mesure par collection electrostatique sur plaque tournante. (author) [Spanish] El autor expone los principios en que se basa la deteccion de las fallas en los revestimientos de los elementos combustibles y describe las caracteristicas principales de la instalacion, que son: a) automatizacion integral, b) centralizacion de las informaciones en la sala de mandos, y c) medicion por recoleccion electrostatica sobre una placa giratoria. (author) [Russian] Izlagayutsya printsipy, na kotorykh osnovano obnaruzhenie razryva obolochki, opisyvaetsya konstruirovanie ustanovok. Osnovnye kharakteristiki takovy: a) integral'nyj avtomatizm, b) tsentralizatsiya informatsii v komandnom zale i c) izmerenie putem ehlektrostaticheskogo sobiraniya na povorachivayushchejsya plastinke. (author)

  1. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1984; Istrazivacki nuklearni reaktor RA - Deo I - Pogon, odrzavanje i eksploatacija nuklearnog reaktora RA u 1984. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1984-12-15

    During the 1984 the reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981. Operation of the primary cooling system was changed in order to avoid appearance of the previously noticed aluminium oxyhydrate on the surface of the fuel element claddings. The new cooling regime enabled more efficient heavy water purification. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks are planned: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. Financing of the planned activities will be partly covered by the IAEA. this Part I of the report includes 8 Annexes describing in detail the reactor operation, and 6 special papers dealing with the problems of reactor operation and utilization.

  2. Development of a methodology for safety classification on a non-reactor nuclear facility illustrated using an specific example; Entwicklung einer Methodik zur Sicherheitsklassifizierung fuer eine kerntechnische Anlage ohne Reaktor an einem spezifischen Beispiel

    Energy Technology Data Exchange (ETDEWEB)

    Scheuermann, F.; Lehradt, O.; Traichel, A. [NUKEM Technologies Engineering Services GmbH, Alzenau (Germany)

    2015-07-01

    To realize the safety of personnel and environment systems and components of nuclear facilities are classified according to their potential danger into safety classes. Based on this classification different demands on the manufacturing quality result. The objective of this work is to present the standardized method developed by NUKEM Technologies Engineering Services for the categorization into the safety classes restricted to Non-reactor nuclear facilities (NRNF). Exemplary the methodology is used on the complex Russian normative system (four safety classes). For NRNF only the lower two safety classes are relevant. The classification into the lowest safety class 4 is accordingly if the maximum resulting dose following from clean-up actions in case of incidents/accidents remains below 20 mSv and the volume activity restrictions of set in NRB-99/2009 are met. The methodology is illustrated using an example. In short the methodology consists of: - Determination of the working time to remove consequences of incidents, - Calculation of the dose resulting from direct radiation and due to inhalation during these works. The application of this methodology avoids over-conservative approaches. As a result some previously higher classified equipment can be classified into the lower safety class.

  3. Report on the interpretation of critical experiments in the Siemens-Argonaut-Reactor Graz to study water ingress into spherical elements. Ergebnisbericht zur Auslegung kritischer Experimente am Siemens-Argonaut-Reaktor Graz zum Studium des Wassereinbruches im Kugelhaufen

    Energy Technology Data Exchange (ETDEWEB)

    Schuerrer, F [Technische Univ., Graz (Austria). Inst. fuer Theoretische Physik und Reaktorphysik; Neef, R D [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung

    1979-04-15

    The experiments described are of interest in the study of water contamination in HTR fuel elements. The Siemens Argonaut Reactor (SAR) has been considered as a research tool for a simulation experiment. Following a brief description of the SAR, planned programs are discussed in 'dry' and 'wet' cores. Detector foil types and locations are noted. A theoretical model is developed and nuclide concentrations estimated in the various spectral zones. Reactivity calculations have been made and are summarised for various H{sub 2}O percentage concentrations. The discussion is supported by simplified core layout diagrams and graphs of core flux distributions. Neutron diffusion and spectra calculations are referenced to computer programs used by KFA-Juelich, published elsewhere, and include GAM, THERMOS, MUPO and EXTERMINATOR-2. (G.C.)

  4. NPP Grafenrheinfeld. No thank you. Questions and answers concerning the oldest operating crack susceptible reactor in the federal republic; AKW Grafenrheinfeld. Nein danke. Fragen und Antworten zum aeltesten noch laufenden und rissanfaelligsten Reaktor der Republik

    Energy Technology Data Exchange (ETDEWEB)

    Darge, Tobias

    2014-03-15

    The brochure discusses questions on the NPP Grafenrheinfeld. concerning the following issues: reactor type, operating company, nuclear fuel, licensing, radioactive waste management, susceptibility to damage, stress test results, critical arguments concerning the reactor pressure vessel steel and the containment, cracks in the primary circuit, safety in case of an aircraft crash, possibility of a severe accident, consequences of a severe accident, medical emergency plan, public information for an emergency case, shutdown at the end of 2015, necessity of a new power plant for energy security?.

  5. RELAP5 SIMULATION FOR SEVERE ACCIDENT ANALYSIS OF RSG-GAS REACTOR

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2018-01-01

    SIMULASI RELAP5 UNTUK ANALISIS KECELAKAAN PARAH PADA REAKTOR RSG-GAS. Reaktor riset di dunia diketahui lebih aman dari pada reaktor daya karena desainnya yang lebih sederhana pada teras dan karakteristika operasinya. Namun demikian, potensi bahaya reaktor riset terhadap publik dan lingkungan tidak bisa diabaikan karena beberapa fitur tertentu. Oleh karena itu, level keselamatan reaktor riset harus jelas ditunjukkan dalam Laporan Analisis Keselamatan (LAK dalam bentuk analisis keselamatan yang dilakukan dengan berbagai macam pendekatan dan metode dan didukung dengan alat komputasi. Tujuan penelitian ini adalah untuk mensimulasikan beberapa kecelakaan parah pada reaktor RSG-GAS yang dapat menyebabkan kerusakan bahan bakar untuk memperkuat hasil analisis kecelakaan parah yang sudah ada dalam LAK. Simulation dilakukan dengan program perhitungan RELAP5/SCDAP/Mod3.4 yang memiliki kemampuan untuk memodelkan elemen bahan bakar tipe pelat di RSG-GAS. Tiga kejadian telah disimulasikan yaitu hilangnya aliran primer dan sekunder dengan kegagalan reaktor untuk dipadamkan, tersumbatnya beberapa kanal pendingin bahan bakar pada daya penuh, dan hilangnya aliran primer dan sekunder yang diikuti dengan tersumbatnya beberapa kanal pendingin bahan bakar setelah reaktor padam. Kejadian pertama akan membahayakan pelat bahan bakar dengan naiknya temperatur kelongsong hingga titik lelehnya yaitu 590 °C. Tersumbatnya satu atau beberapa kanal pada satu elemen bahan bakar menyebabkan konsekuensi yang berbeda pada pelat bahan bakar, dimana paling sedikit tersumbatnya 2 kanal akan merusak satu pelat bahan bakar, apalagi tersumbatnya satu elemen bahan bakar. Kombinasi antara hilangnya aliran pendingin primer dan sekunder yang diikuti dengan tersumbatnya satu kanal bahan bakar setelah reaktor dipadamkan menyebabkan naiknya temperatur kelongsong di bawah titik lelehnya yang berarti sirkulasi alam yang terbentuk dan daya yang terus turun cukup untuk mendinginkan elemen bahan bakar. Kata kunci

  6. Põhja-Korea lubas taas jätta tuumaplaanid / Jürgen Tamme

    Index Scriptorium Estoniae

    Tamme, Jürgen

    2007-01-01

    Pekingis toimunud kuuepoolsete desarmeerimiskõneluste tulemusena on Põhja-Korea nõus andma ülevaate riigi tuumarajatistest ning alustama nende likvideerimist. 14. juulil suleti Yongbyoni tuumakeskuse reaktor. Lisa: Tuumavaidlus

  7. KARAKTERISTIK REAKTIVITAS TERAS KERJA RSG-GAS SELAMA 30 TAHUN BEROPERASI

    Directory of Open Access Journals (Sweden)

    Tukiran Surbakti

    2017-06-01

    Full Text Available RSG-GAS, mulai dari komisioning, operasi teras kerja hingga kini telah 30 tahun beroperasi sehingga perlu dilakukan evaluasi keselamatan parameter neutroniknya. Untuk tujuan keselamatan telah dilakukan berbagai aktivitas penelitian, baik yang berhubungan dengan operasi, keselamatan, maupun dalam rangka penggunaan reaktor. Analisis dan pengelolaan besaran reaktivitas yang menunjang keselamatan operasi reaktor sangat penting dilakukan karena besaran ini mempengaruhi desain, kendali dan jadual operasi reaktor. Besaran tersebut dapat ditentukan melalui pengukuran reaktivitas batang kendali dan eksperimen pemuatan bahan bakar di dalam teras. Pengukuran reaktivitas batang kendali yang dilakukan pada setiap awal siklus teras (dengan kondisi teras dingin dan bersih, bebas pengaruh xenon, menghasilkan nilai reaktivitas batang kendali yang dapat digunakan untuk menentukan nilai reaktivitas lainnya seperti reaktivitas lebih, reaktivitas padam dan reaktivitas total. Pengelolaan reaktivitas teras telah dilakukan dengan baik selama 30 tahun dalam rangka mendukung operasi reaktor untuk keperluan penelitian dan iradiasi target.

  8. Adamkus : EU might be willing to negotiate on postponing Ignalina reactor closure

    Index Scriptorium Estoniae

    2008-01-01

    President Valdas Adamkus teatas võimalusest, et EL alustab arutelu Ignalina tuumaelektrijaama sulgemise edasilükkamisest. Uue jaama valmimise aeg pole teada ning praegune reaktor on tänu rootslaste abile turvaline

  9. Studi Awal Desain Pebble Bed Reactor Berbasis Htr-pm Dengan Skema Resirkulasi Bahan Bakar Once-through-then-out

    OpenAIRE

    Setiadipura, Topan; Pane, Jupiter Sitorus; Zuhair, Zuhair

    2016-01-01

    STUDI AWAL DESAIN PEBBLE BED REACTOR BERBASIS HTR-PM DENGAN RESIRKULASI BAHAN BAKAR ONCE-THROUGH-THEN-OUT. Reaktor nuklir tipe pebble bed reactor (PBR) adalah salah satu reaktor canggih dengan fitur keselamatan pasif yang kuat. Pada desain tipe ini berpotensi untuk dilakukan kogenerasi yang bermanfaat untuk pengolahan berbagai mineral di berbagai pulau di Indonesia. Operasi PBR dapat lebih disederhanakan dengan menerapkan skema pengisian bahan bakar once-through-then-out (OTTO) dimana bahan b...

  10. Efek Durasi Pencahayaan Pada Sistem HRAR Untuk Menurunkan Kandungan Minyak Solar Dalam Air Limbah

    Directory of Open Access Journals (Sweden)

    Dian Puspitasari

    2014-09-01

    Full Text Available Kandungan minyak di dalam air limbah industri perminyakan umumnya bersifat toksik terhadap mikroorganisme dan mengganggu proses pengolahan secara biologis. Sistem HRAR diperkirakan dapat mengatasi hambatan tersebut melalui proses fotosintesis untuk menghasilkan oksigen yang dibutuhkan mikroorganisme dalam mendegradasi senyawa hidrokarbon. Penelitian ini bertujuan mengkaji pengaruh perpanjangan waktu pencahayaan pada kemampuan HRAR dalam menurunkan kandungan minyak di dalam limbah. Variabel yang digunakan pada penelitian ini adalah variasi durasi pencahayaan dan variasi penambahan volume minyak solar yang ditambahkan ke dalam reaktor. Variasi durasi pencahayaan yang digunakan adalah pencahayaan selama 12 jam dan pencahayaan selama 24 jam. Sedangkan penambahan volume minyak solar ke dalam masing-masing reaktor adalah sebesar 346 ppm, 519 ppm dan 692 ppm. Hasil yang didapatkan dari penelitian ini adalah durasi pencahayaan selama 12 jam memiliki efek yang lebih baik terhadap penurunan konsentrasi minyak dibandingkan pencahayaan selama 24 jam. Hal ini dapat terlihat dari baiknya pertumbuhan alga dan bakteri di dalam reaktor serta tingginya penurunan konsentrasi minyak solar di dalamnya. Penurunan konsentrasi minyak solar terbaik terdapat pada reaktor dengan penambahan minyak solar sebesar 346 ppm. Pada reaktor dengan durasi pencahayaan selama 12 jam terjadi penurunan konsentrasi minyak sebesar 78,4%. Sedangkan penurunan kandungan minyak solar pada reaktor dengan durasi pencahayaan selama 24 jam adalah sebesar 73,9%.

  11. SAFIRA. Sub-project B 1.3: Development of coupled in-situ reactors and optimisation of the geochemical processes in the discharge of different in situ reactor sytems. Final report; SAFIRA. Teilprojekt B 1.3: Entwicklung von gekoppelten in situ-Reaktoren und Optimierung der geochemischen Prozesse im Abstrom von verschiedenen in situ-Reaktor-Systemen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Dahmke, A.; Schaefer, D.; Koeber, R.; Plagentz, V.

    2002-12-01

    The Bitterfeld ground water is contaminated with many different pollutants over a large area. Long-term measures like reactive barriers for purification are required. However, groundwater contaminated with multiple contaminants cannot be purified by a single reactive material; for this reason, the effectivity of combinations of different reactive materials was investigated. Of the combinations investigated, reducing iron and activated carbon connected in series was the most effective: The iron will remove the reducible chlorinated hydrocarbons, while the rest of the contaminants are adsorbed to the activated carbon. Iron and ORC was another interesting option, but the combination of iron and activated carbon was found to be the most favourable option. Until a better method is available, it is recommended to connect iron and activated carbon in parallel for removing contaminant mixtures. Directly behind reactive iron barriers (also when combined with activated carbon), the limiting values of the Freshwater Ordinance for Fe(II) and pH are exceeded. Directly behind ORC reactors, the limiting values for Mg and pH are exceeded. Investigations in the outflow of these reactive materials showed that the high pH values are buffered by contact with the aquifer material to values typical of aquifers, which usually are below the limiting values of the Freshwater Ordinance. However, as the buffer capacity of the soil is exhausted, a zone with a higher pH starts to grow in the aquifer. The growth of this zone depends on the pH and on the aquifer material. Especially in soils as found at Bitterfeld, with a high concentration of organic matter, we find long-term desorption of pollutants from the aquifer materials which will burden the purified water leaving the water treatment system and prohibit its utilization. [German] Der Grundwasserleiter im Raum Bitterfeld ist grossraeumig mit vielen verschiedenen Substanzen kontaminiert. Aufgrund der grossraeumigen Erstreckung kommen nur langfristig kostenguenstige passive Massnahmen wie reaktive Barrieren zur Sanierung in Frage. Grundwasser, das mit mehreren und unterschiedlich reagierenden Stoffen kontaminiert ist, kann jedoch nicht mit Hilfe eines einzelnen reaktiven Materials gereinigt werden, daher wurde die Effektivitaet von Kombinationen unterschiedlicher reaktiver Materialien zur Sanierung untersucht. Von den untersuchten Kombinationen erwies sich die Hintereinanderschaltung von reduzierendem Eisen und Aktivkohle als besonders effektiv. Reduzierbare chlorierte Kohlenwasserstoffe werden im Eisen entfernt, die verbleibenden Kontaminanten adsorbieren auf der Aktivkohle. Auch die Hintereinanderschaltung von Eisen und Sauerstoff abgebenden ORC, in denen ein aerober mikrobieller Abbau statt findet, ist zur Entfernung von Mischkontaminationen geeignet. Eine Kostenschaetzung zeigt, dass die Kombination von Eisen und Aktivkohle in Abhaengigkeit von der Zusammensetzung der Kontamination guenstiger als Aktivkohle allein sein kann und generell guenstiger als die Kombination von Eisen und ORC ist. Ohne ein guenstigeres Verfahren zum Einbringen von Sauerstoff in den Aquifer wird die Hintereinanderschaltung von Eisen und Aktivkohle zur Sanierung von Mischkontaminationen empfohlen. Im direkten Abstrom von reaktiven Eisenbarrieren (auch in Kombination mit Aktivkohle) sind die Grenzwerte fuer Fe(II) und pH entsprechend der Trinkwasserverordnung ueberschritten. Im Abstrom von ORC-Reaktoren werden die zulaessigen Werte fuer Mg und pH ueberschritten. Untersuchungen im Abstrom dieser reaktiven Materialien zeigen, dass die hohen pH-Werte durch den Kontakt mit dem Aquifermaterial auf Aquifer-typische Werte gepuffert werden, die ueblicherweise unter den Grenzwerten der Trinkwasserverordnung liegen. Mit Erschoepfen der Pufferkapazitaet des Bodens breitet sich jedoch eine Zone mit erhoehtem pH-Wert im Aquifer aus. Die Geschwindigkeit dieser Ausbreitung haengt vom pH-Wert und dem Aquifermaterial ab. Gerade fuer sehr Organik reiche Boeden wie in Bitterfeld wird auch eine lange anhaltende Desorption von Schadstoffen aus dem Aquifermaterial beobachtet, durch die das saubere Wasser, das die Sanierungsanlage verlaesst, wieder mit Schadstoffen beladen wird, was zu einer anhaltenden Nutzungseinschraenkung fuehrt. (orig.)

  12. Radioactivity in the environment of the RA nuclear reactor in Vinca for the period 1977-1980. Material prepared for the RA reactor safety report; Radioaktivnost okoline nuklearnog reaktora RA u Vinci u periodu 1977-1980, Materijal pripremljen za izradu Sigurnosnog izvestaja za reaktor RA

    Energy Technology Data Exchange (ETDEWEB)

    Ajdacic, N; Martinc, R [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1980-12-15

    Review of the environmental monitoring data presented in this report is prepared for the RA reactor safety report. These data resulted from four-year monitoring of precipitations and deposited dust. Measurements were done daily. In addition to these data, tables contain mean daily values, total monthly values of beta activities of precipitation from 1977 - 1980. Radioactivity control of the RA reactor environment showed that there was no significant discrepancy compared to the mean values for several years, apart from seasonal variations and meteorological influences. In the period from October 1976 to mid 1978 a number of higher values were recorded probably due to nuclear explosions. During 1979 the general activity level was relatively low, showing increase tendency during 1980.

  13. THERMAL-HYDRAULIC ANALYSIS OF SMR WITH NATURALLY CIRCULATING PRIMARY SYSTEM DURING LOSS OF FEED WATER ACCIDENT

    Directory of Open Access Journals (Sweden)

    Susyadi Susyadi

    2016-09-01

    ABSTRAK Reaktor daya kecil modular (SMR memiliki beberapa keunggulan dibanding reaktor daya besar konvensional. Dengan disain yang lebih sederhana dan terintegrasi, penerapan hukum alamiah untuk sistem keselamatannya dan biaya modal yang rendah, reaktor ini sangat cocok untuk dibangun di Indonesia. Salah satunya disain SMR yang sedang dikembangkan menerapkan gaya penggerak alami untuk sistim pendingin primernya. Dengan disain seperti itu, adalah sangat penting untuk memahami implikasinya terhadap aspek keselamatan pada seluruh kondisi operasi. Salah satu yang perlu diinvestigasi adalah kecelakaan kehilangan air umpan (LoFW. Pada studi ini, dilakukan analisis kinerja thermal hidrolik SMR yang menggunakan sistim pendinginan primer sirkulasi alam saat kecelakaan LoFW. Tujuannya adalah untuk menginvestigasi karakteristik aliran sistem primer saat kecelakaan LoFW dan untuk memastikan apakah aliran sirkulasi alam cukup untuk memindahkan panas dari teras guna menjaga kondisi tetap aman selama kecelakaan tersebut. Metoda yang digunakan adalah dengan merepresentasikan sistem reaktor ke dalam model-model generik program RELAP5 dan melakukan simulasi numerik. Hasil perhitungan menunjukkan bahwa setelah kejadian pemicu dan trip reaktor, pada sisi primer laju alirnya berfluktuasi secara signifikan dan temperatur pendinginnya menurun secara bertahap sedangkan  pada sisi sekunder kondisi uap berubah menjadi uap jenuh. Laju alir turun dari ~711 kg/detik menjadi ~263 kg/detik sebelum kembali naik lagi pada t=~46 detik. Saat laju alir di titik terendah, temperatur pusat bahan bakar dan fluida pendingin adalah sekitar  ~565 K dan  ~554 K, yang menujukkan bahwa temperatur bahan bakar masih jauh di bawah batas disain dan temperatur fluidanya juga berada di bawah titik saturasi. Keadaan ini menunjukkan bahwa saat transien kedua parameter utama termohidrolik reaktor tetap dalam kondisi yang dapat diterima sehingga dapat disimpulkan  bahwa saat  kecelakaan kehilangan air umpan, SMR

  14. The design, development and operation of a compact nuclear power plant simulator

    International Nuclear Information System (INIS)

    Lynch, M.F.

    1987-01-01

    This paper discusses the philosophy and technological considerations necessary for constructing and utilizing a plant specific compact nuclear power plant simulator, how it compares to full scope replica simulators, engineering simulators, part task simulators and basic principles training simulators. Included in this discussion are the design process, scope of simulation, the manufacturing process, test programs and experiences with operator training. Items addressed include the applicability and use of a compact simulator, how well it reproduces the actual reference plant, how well the transferral of knowledge is accomplished and what financial considerations need to be evaluated. This paper tries to provide the details on just how this type of machine was designed and developed by Westinghouse for the Swiss Utility, Nordostschweizerische Kraftwerke (NOK) AG

  15. Perancangan dan Simulasi MRAC untui Proses Pengendalian Temperatur pada Continuous Stirred Tank Reactor (CSTR

    Directory of Open Access Journals (Sweden)

    Amelia Sylvia

    2014-03-01

    Full Text Available Temperatur merupakan salah satu variabel proses dasar yang dikendalikan untuk menjaga suhu cairan di dalam reaktor. Model Reference Adaptive Controller (MRAC dengan MIT rule dipilih untuk mencapai spesifikasi respon yang diinginkan pada CSTR. Beban yang bervariasi berupa debit aliran likuid yang masuk ke dalam reaktor dapat menyebabkan perubahan parameter yang mempengaruhi perubahan temperatur output produk pada CSTR. Sebuah simulasi dilakukan dengan menggunakan MATLAB dan hasilnya dianalisa. Respon plant dapat melakukan adaptasi parameter – parameter kontrolernya cukup baik pada nilai gain adaptasi dengan rentang 0.00000010000 sampai 0.00000000001. Waktu yang dibutuhkan untuk mengatasi beban yang bervariasi berupa debit aliran yang masuk ke dalam reaktor dengan nilai yang maksimal (1.5 m^3/min menghasilkan respon plant lebih cepat 42 detik dari pada debit aliran masuk dengan nilai yang nominal (1 m^3/min 63 detik dan minimal (0.5 m^3/min 75 detik.

  16. Studi Efisiensi Penyisihan COD dalam Lindi dengan Sistem Evapotranspirasi Menggunakan Tumbuhan Sente (Alocasia macrorrhiza dan Rumput Belulang (Eleusine indica

    Directory of Open Access Journals (Sweden)

    Badrus Zaman

    2017-11-01

    Full Text Available COD dalam lindi merupakan salah satu parameter yang secara umum berada pada konsentrasi yang tiggi sebagai salah satu hasil biodegradasi material organik dan anorganik dalam sampah di TPA. Sistem evapotranspirasi yang menggunakan tumbuhan lokal merupakan salah satu sistem yang menjanjikan. Penelitian dilakukan untuk mengetahui efisiensi penyisihan COD dalam lindi dengan reaktor evapotranspirasi secara kontinyu yang menggunakan tumbuhan Tumbuhan Sente (Alocasia macrorrhiza dan Rumput Belulang (Eleusine indica. Hasil uji menunjukkan efisiensi pada semua reaktor mulai sekitar hari ke 3 hingga hari ke 25 mengalami fluktuasi yang cenderung menurun (dari ± 75% menjadi ± 50%, tetapi hari selanjutnya cenderung meningkat. Pola tersebut dipengaruhi oleh peran media tanam, bakteri dalam media tanam, bakteri pada akar tumbuhan dan aktivitas metabolisme tumbuhan uji. Secara keseluruhan reaktor yang menggunakan Tumbuhan Sente (Alocasia macrorrhiza lebih fluktuatif dibandingkan denga menggunakan Rumput Belulang (Eleusine indica yang dipengaruhi pola pertumbuhan dan perkembangannya.

  17. Integrating 3D CAD data for manufacturing and fabrication the core model of reactor TRIGA PUSPATI

    International Nuclear Information System (INIS)

    Abu Bakar Harun

    2005-01-01

    This paper describe the intrigue integration of digital 3 Dimensional Computer Aided Design (3D CAD) data manipulation for the Core Model fabrication of REAKTOR TRIGA PUSPATI and ready for mass manufacturing. 3 Dimensional CAD data from Computer Aided Design program will be used as an interpreter in the fabrication of this project. The Core Model of REAKTOR TRIGA PUSPATI will be fabricated with the aid of 3D CAD drawings and digital files. The components will be segregated and divided into 2 categories namely Conventional d Rapid Fabrication. (Author)

  18. Services provided to other organizational units, Annex 7; Prilog 7 - Usluge drugim organizacionim jedinicama

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-12-15

    During 1965, the RA reactor staff provided services to other organizational units in: designing new systems, repair of electronic equipment, installing measuring devices, constructing mechanical elements in the workshop. [Serbo-Croat] U toku 1965. godine reaktor RA je pruzio usluge drugim jedinicama i to: projektovanjem novih sistema, popravkom elektronskih uredjaja, instaliranjem mernih sistema, izradom mehanickih elemenata i sklopova u mehanickoj radionici.

  19. Ultrasonics aids the identification of failed fuel rods

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    Over a number of years Brown Boveri Reaktor of West Germany has developed and commercialized an ultrasonic failed fuel rod detection system. Sipping has up to now been the standard technique for failed fuel detection, but sipping can only indicate whether or not an assembly contains defective rods; the BBR system can tell which rod is defective. (author)

  20. Eesti Energia valib Ameerikas tuumajaamale reaktoreid / Andres Reimer

    Index Scriptorium Estoniae

    Reimer, Andres

    2009-01-01

    Eesti Energia liitus rahvusvahelise tuumareaktoriprojektiga IRIS, projekti eesmärk on luua lihtsa konstruktsiooniga, ohutu ja suhteliselt odav reaktor. Jürgen Ligi sõnul saab Eesti oma IRIS-tüüpi tuumajaama juba 2019. aastal. Joonis: Potentsiaalne tuumajaam

  1. Na velikosti nezáleží - je to fraktální

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2014-01-01

    Roč. 7, červen (2014) Institutional support: RVO:61389021 Keywords : fusion * tokamak * fractal * turbulence * Bohm diffusion Subject RIV: BL - Plasma and Gas Discharge Physics http://3pol.cz/1614-na-velikosti-nezalezi-je-to-fraktalni-aneb-slunecni-termojaderny-reaktor

  2. Pengolahan Sampah Secara Pirolisis dengan Variasi Rasio Komposisi Sampah dan Jenis Plastik

    Directory of Open Access Journals (Sweden)

    Qonita Rachmawati

    2015-03-01

    Full Text Available Pada tahun 2012, sampah yang dihasilkan sebesar 1200 ton/hari. Jika permasalahan sampah di Kota Surabaya tidak ditangani dengan baik maka akan menimbulkan beberapa masalah antara lain: masalah kesehatan dan masalah kebersihan. Oleh karena itu, diperlukan metode yang dapat mengolah sampah namun tidak menimbulkan masalah baru lainnya. Salah satu metode pengolahan sampah yang telah dikembangkan, yaitu metode pirolisis. Penelitian ini bertujuan untuk menentukan pengaruh jenis plastik dan komposisi terhadap produk hasil pirolisis. Pada penelitian ini digunakan reaktor dengan kapasitas 500 g yang berbahan stainless steel. Variabel yang digunakan yaitu jenis sampah plastik dan komposisi sampah. Jenis sampah yang digunakan yaitu sampah plastik HDPE (High Density Polyethylene, PET (Poly Ethylene Terephthalate, dan PS (Poly Styrene. Komposisi sampah yang digunakan antara lain: 100:0, 75:25, dan 50:50. Temperatur yang digunakan pada reaktor yaitu 500°C dengan waktu 30 menit. Penelitian dimulai dari persiapan bahan uji, persiapan reaktor, percobaan pendahuluan, dan penelitian dengan reaktor pirolisis. Selanjutnya dilakukan analisis untuk masing-masing hasil produk. Penelitian ini jenis sampah plastik yang menghasilkan gas tertinggi yaitu jenis plastik PET sebesar 45,40% dan jenis plastik yang menghasilkan wax tertinggi yaitu jenis plastik HDPE sebesar 69,91%. Sedangkan komposisi yang menghasilkan gas tertinggi yaitu komposisi dengan ranting 25% dan PET 75% sebesar 71,24% dan komposisi yang menghasilkan wax tertinggi yaitu komposisi dengan ranting 25% dan PS 75% sebesar 61,36%.

  3. Servicing NPPs in Japan, Korea and Taiwan

    International Nuclear Information System (INIS)

    Bohmann, W.; Poetz, F.

    1991-01-01

    More than 45 comparable orders (for equipment or services ) have been carried out in Japan, Korea and Taiwan by ABB Reaktor since 1982. Recently contracts to deliver inspection and repair equipment for the core baffle former bolts of Japanese NPPs, and in 1990, a contract to clean two steam generators in Korea, together with Pacific Nuclear Services, USA, were won. 2 figs

  4. ANALYSIS OF GAMMA HEATING AT TRIGA MARK REACTOR CORE BANDUNG USING PLATE TYPE FUEL

    Directory of Open Access Journals (Sweden)

    Setiyanto Setiyanto

    2016-10-01

    Full Text Available ABSTRACT In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities and central irradiation position (CIP, especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0,87 W/g, but very low value for Lazy Susan position (lest then 0,11 W/g. Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. Keywords: gamma heating, nuclear reactor, research reactor, reactor safety.   ABSTRAK Dengan dihentikannya produksi elemen bakar reaktor jenis Triga oleh produsen, maka semua reaktor TRIGA di dunia terganggu operasinya, termasuk juga reaktor TRIGA 2000 di Bandung. Untuk mendukung pengoperasian reaktor TRIGA Bandung

  5. High-Temperature Gas-Cooled Reactor Critical Experiment and its Application to Thorium Absorption Rates; Experience Critique pour l'Etude d'un Reacteur a Haute Temperature, Refroidi par un Gaz et son Application a la Determination des Taux d'Absorption du Thorium; Kriticheskij opyt, postavlennyj na vysokotemperaturnom reaktore s gazovym okhlazhdeniem, i primenenie ego dlya opredeleniya stepeni pogloshcheniya toriya; Experimento Critico Efectuado en un Reactor de Elevada Temperatura Refrigerado por Gas y su Aplicacion para Calcular los Indices de Absorcion del Torio

    Energy Technology Data Exchange (ETDEWEB)

    Bardes, R. G.; Brown, J. R.; Drake, M. K.; Fischer, P. U.; Pound, D. C.; Sampson, J. B.; Stewart, H. B. [General Dynamics Corporation,San Diego, CA (United States)

    1964-04-15

    In developing the concept of the HTGR and its first prototype at Peach Bottom, General Atomic made the decision that a critical experiment was required to provide adequately certain necessary input data for the nuclear analysis. The specific needs of the nuclear design theory for input data relating to thorium absorptions led to an experimental design consisting of a central lattice-type critical assembly with surrounding buffer and driver regions. This type of assembly, in which the spectrum of interest can be established in the relatively small central lattice having a desired geometry, provides a useful tool for obtaining a variety of input data for nuclear analysis surveys of new concepts. The particular advantages of this approach over that of constructing a mock-up assembly will be discussed, as well as the role of the theory in determining what experiments are most useful and how these experiments are then used in verifying design techniques. Two relatively new techniques were developed for use in the lattice assembly. These were a reactivity oscillation technique for determining the thorium Doppler coefficient, and an activation technique for determining both the resonance integral of thorium dispersed in graphite and its temperature dependence (activation Doppler coefficient). The Doppler coefficient measurement by reactivity oscillation utilized the entire central fuel element in a technique which permitted heating this fuel element to 800 Degree-Sign F and accurately subtracting experimentally the thermal-base effects, that is, those effects not contributing to the thorium resonance capture. Comparison of results with theory for a range of conditions shows excellent agreement. The measurement of the thorium resonance integral and its temperature dependence will be described. The technique developed for measuring resonance capture makes use of gold as the standard and vanadium as die material giving the 1/v absorption rate. This technique is dictated by the fact that the thorium is dispersed in graphite and the usual cadmium-ratio technique is difficult to apply. Comparison of experimental and theoretical results shows excellent agreement over a range of variables. In addition, the results of both activation and reactivity measurements of Doppler coefficient are in agreement, a fact which is felt to be significant in view of the disparity between results from these two techniques in the literature. (author) [French] Lors de l'etude du reacteur HTGR a haute temperature refroidi par un gaz, et de son premier prototype a Peach Bottom, la General Atomic Division de la societe General Dynamics a decide qu'il fallait proceder a une experience critique pour obtenir certaines donnees d'entree necessaires pour l'analyse nucleaire. Aux fins de l'etude nucleaire theorique, les besoins particuliers en donnees d'entree relatives aux absorptions par le thorium ont amene les ingenieurs a concevoir un assemblage experimental critique compose d'un reseau central entoure d*une region tampon et d'une region de commande. Ce type.d'assemblage, dans lequel on peut creer le spectre a mesurer dans le reseau central relativement petit ayant la geometrie voulue, permet d'obtenir des donnees d'entree tres diverses pour les etudes de projets nouveaux, au point de vue de l'analyse nucleaire. Le memoire indique les avantages particuliers que presente cette methode par rapport a celle qiu consiste a construire une maquette, ainsi que le role de la theorie pour determiner quelles experiences sont le plus utiles et comment utiliser ensuite ces experiences dans la verification des procedes d'etude. Les auteurs ont mis au point deux methodes relativement nouvelles qui peuvent etre utilisees avec l'assemblage decrit ci-dessus: une methode d'oscillation de la reactivite pour determiner le coefficient Doppler pour le thorium; une methode d'activation pour determiner a la fois l'integrale de resonance pour le thorium disperse dans le graphite et ses variations en fonction de la temperature (coefficient Doppler d'activation). Pour mesurer le coefficient Doppler par oscillation de la reactivite, on se sert de la totalite de la cartouche centrale au cours d'une operation qui permet de la porter a une temperature pouvant atteindre 425 Degree-Sign C et d'eliminer experimentalement les effets qui ne contribuent pas a la capture de neutrons de resonance par le thorium. Pour toute une gamme d'experiences, les resultats obtenus concordent bien avec les resultats theoriques. Le memoire decrit la mesure de l'integrale de resonance pour le thorium et ses variations en fonction de la temperature. Dans le procede que l'on a mis au point pour mesurer la capture de resonance, on utilise l'or comme etalon et le vanadium comme matiere donnant le taux d'absorption en 1/v. Ce procede a ete choisi parce que le thorium est disperse dans le graphite et qu'il est difficile d'appliquer le procede habituel du rapport cadmium. Les resultats empiriques concordent bien avec les resultats theoriques dans une large gamme de variables. En outre, les resultats des mesures du coefficient Doppler par les deux methodes (oscillation de la reactivite et activation) concordent. Les auteurs estiment que ce fait merite d'etre releve, car dans les ouvrages publies jusqu'ici ces deux procedes donnaient des resultats differents. (author) [Spanish] Al definir los principios teoricos del reactor de elevada temperatura refrigerado por gas, y de su primer prototipo en Peach Bottom, la General Atomic decidio efectuar un experimento critico con miras a reunir ciertos datos de entrada requeridos para el analisis nuclear. Debido a las necesidades especificas de la teoria de las construcciones nucleares, en lo que atane a los datos de entrada acerca de la absorcion del torio, los autores elaboraron un sistema experimental formado por un conjunto critico con reticulado central rodeado por una region amortiguadora y otra activado'. Este tipo de conjunto, en el que el espectro que se ha de analizar puede limitarse al reticulado central relativamente pequeno, y cuya geometria puede determinarse a voluntad, permite obtener diversos datos de entrada para los estudios relativos a analisis nucleares de instalaciones nuevas. Los autores describen las ventajas de este metodo en comparacion con el de una maqueta y la funcion de la teoria, consistente en determinar cuales son los experimentos mas utiles y la manera en que deben utilizarse para comprobar los proyectos. Los autores elaboraron dos tecnicas relativamente nuevas para su utilizacion en un conjunto dotado de reticulado. Una se basa en la tecnica de oscilacion de la reactividad para determinar el coeficiente Doppler correspondiente al torio, y la otra en la activacion para determinar la integral de resonancia del torio dispersado en el grafito y su variacion en funcion de la temperatura (coeficiente Doppler de activacion). Para medir el coeficiente Doppler por oscilacion de la reactividad, se utilizo todo el elemento combustible situado en el centro, calentando este elemento hasta 800 Degree-Sign F y procediendo a una sustraccion experimental exacta de los efectos que no contribuyen a la captura por resonancia en el torio. Los resultados obtenidos concuerdan satisfactoriamente con los valores teoricos en diversas condiciones. Los autores describen la medicion de la integral de resonancia en el torio y su variacion en funcion de la temperatura. En la tecnica que han elaborado para medir la captura por resonancia, se utiliza oro como patron y vanadio como el material que cumple la ley 1/v. La eleccion de esta tecnica obedece al hecho de que el torio se dispersa en el grafito y a la dificultad de aplicar el metodo basado en la razon cadmica. Los resultadoe experimentales concuerdan con los teoricos en diversas condiciones. Asimismo, concuerdan entre si las mediciones del efecto Doppler efectuadas por activacion y la realizadas por determinacion de la reactividad, lo que se considera significativo debido a que en las obras publicadas subsiste cierta divergencia en los resultados obtenidos con ambas tecnicas. (author) [Russian] Pri razrabotke idei reaktora HTGR i ego pervogo prototipa v Pich- Bottome prishli k resheniju o neobhodimosti obespechit' sootvetstvujushhie ishodnye dannye dlja provedenija jadernogo analiza. Konkretnye potrebnosti teorii jadernogo proektirovanija na ishodnyh dannyh otnositel'no pogloshhenija torija priveli k sozdaniju jeksperimental'nogo proekta, sostojashhego iz kriticheskoj sborki tipa sborki s central'noj reshetkoj s okruzhajushhim ammortizatorom i peredvizhnymi aktivnymi zonami. Sborka takogo tipa, v kotoroj v sravnitel'no nebol'shoj central'noj reshetke s zhelaemoj geometriej mozhet byt' ustanovlen predstavljajushhij interes spektr, javljaetsja poleznoj ustanovkoj dlja poluchenija raznoobraznyh ishodnyh dannyh v celjah provedenija jadernogo analiza novyh idej. Obsuzhdajutsja konkretnye preimushhestva jetogo metoda po sravneniju so stroitel'stvom sborki-modeli, a takzhe rol' teorii v opredelenii, kakie opyty javljajutsja naibolee poleznymi i kak jeti opyty zatem ispol'zujutsja pri proverke metodov proektirovanija. - Byli razrabotany dva sravnitel'no novyh metoda dlja ispol'zovanija v sborke reshetok - metod izmerenija kolebanij reaktivnosti dlja opredelenija kojefficienta Dopplera dlja torija i metod aktivacii dlja opredelenija kak rezonansnogo integrala torija, dispergirovannogo v grafite, tak i ego zavisimosti ot temperatury (kojefficient aktivacii Dopplera). Pri izmerenii kojefficienta Dopplera putem opredelenija kolebanij reaktivnosti ves' central'nyj toplivnyj jelement ispol'zovalsja takim obrazom, chto byla vozmozhnost' osushhestvit' nagrev toplivnogo jelementa do 800 Registered-Sign F i tochno opredelit' opytnym putem teplovye jeffekty, t.e. te jeffekty, kotorye ne okazyvajut vlijanija na velichinu rezonansvogo zahvata torija. Sravnenie rezul'tatov s teoriej dlja rjada uslovij svidetel'stvuet o prekrasnom soglasovanii. Daetsja opisanie izmerenija rezonansnogo integrala torija i ego zavisimosti ot temperatury. Pri jetom metode dlja izmerenija rezonansnogo zahvata v kachestve standarta ispol'zuetsja zoloto, a v kachestve materiala, dajushhego velichinu pogloshhenija, podchinjajushhujusja zakonu 1/v, -vanadij. Jetot metod obuslovlivaetsja tem faktom, chto torij dispergiruetsja v grafite i trudno primenjat' obychnyj metod kadmievogo otnoshenija. Sravnenie jeksperimental'nyh i teoreticheskih rezul'tatov svidetel'stvuet o prekrasnom soglasvi vo vsem diapazone peremennyh.. Krome togo, soglasujutsja rezul'taty izmerenija kojefficienta Dopplera kak metodom aktivacii, tak i metodom opredelenija reaktivnosti,-fakt, kotoryj, kak polagajut, javljaetsja vazhnym vvidu nalichija rashozhdenija mezhdu rezul'tatami primenenija jetih dvuh metodov, imejushhimisja v literature. (author)

  6. Simulator testing of the Westinghouse aware alarm management system

    Energy Technology Data Exchange (ETDEWEB)

    Carrera, J P; Easter, J R; Roth, E M [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1997-09-01

    Over the last year, Westinghouse engineers and operators from the Beznau nuclear power station (KKB), owned by the Nordostschweizerische Krafwerke AG of Baden, Switzerland, have been installing and testing the Westinghouse AWARE Alarm Management System in Beznau/SNUPPS operator training simulator, owned and operated by the Westinghouse Electric Corp., in Waltz Mill, PA, USA. The testing has focused primarily on validating the trigger logic data base and on familiarizing the utility`s training department with the operation of the system in a real-time environment. Some of the tests have included plant process scenarios in which the computerized Emergency Procedures were available and used through the COMPRO (COMputerized PROcedures) System in conjunction with the AWARE System. While the results to date are qualitative from the perspective of system performance and improvement in message presentation, the tests have generally confirmed the expectations of the design. There is a large reduction in the number of messages that the control room staff must deal with during major process abnormalities, yet at times of relative minor disturbances, some additional messages are available which add clarification, e.g., ``Pump Trouble`` messages. The ``flow`` of an abnormality as it progresses from one part of the plant`s processes to another is quite visible. Timing of the messages and the lack of message avalanching is proving to give the operators additional time to respond to messages. Generally, the anxiety level to ``do something`` immediately upon a reactor trip appears to be reduced. (author). 8 refs.

  7. Simulator testing of the Westinghouse aware alarm management system

    International Nuclear Information System (INIS)

    Carrera, J.P.; Easter, J.R.; Roth, E.M.

    1997-01-01

    Over the last year, Westinghouse engineers and operators from the Beznau nuclear power station (KKB), owned by the Nordostschweizerische Krafwerke AG of Baden, Switzerland, have been installing and testing the Westinghouse AWARE Alarm Management System in Beznau/SNUPPS operator training simulator, owned and operated by the Westinghouse Electric Corp., in Waltz Mill, PA, USA. The testing has focused primarily on validating the trigger logic data base and on familiarizing the utility's training department with the operation of the system in a real-time environment. Some of the tests have included plant process scenarios in which the computerized Emergency Procedures were available and used through the COMPRO (COMputerized PROcedures) System in conjunction with the AWARE System. While the results to date are qualitative from the perspective of system performance and improvement in message presentation, the tests have generally confirmed the expectations of the design. There is a large reduction in the number of messages that the control room staff must deal with during major process abnormalities, yet at times of relative minor disturbances, some additional messages are available which add clarification, e.g., ''Pump Trouble'' messages. The ''flow'' of an abnormality as it progresses from one part of the plant's processes to another is quite visible. Timing of the messages and the lack of message avalanching is proving to give the operators additional time to respond to messages. Generally, the anxiety level to ''do something'' immediately upon a reactor trip appears to be reduced. (author). 8 refs

  8. EFEK DENSITAS BAHAN BAKAR TERHADAP PARAMETER KOEFISIEN REAKTIVITAS TERAS RRI

    Directory of Open Access Journals (Sweden)

    Rokhmadi Rokhmadi

    2015-03-01

    Full Text Available Manfaat yang luas penggunaan reaktor riset membuat banyak negara membangun reaktor riset baru. Kecenderungan saat ini adalah tipe reaktor serbaguna (MPR dengan teras yang kompak untuk mendapatkan fluks neutron yang tinggi dengan daya yang relatif rendah. Reaktor riset yang ada di Indonesia usianya sudah tua semuanya. Oleh karena itu diperlukan desain reaktor riset baru sebagai alternatif, disebut reaktor riset inovatif (RRI, kelak pengganti reaktor riset yang sudah ada. Tujuan dari riset ini untuk melengkapi data desain RRI sebagai salah satu persyaratan untuk perizinan desain. Perhitungan dilakukan untuk memperoleh nilai koefisien reaktivitas teras RRI dengan konfigurasi teras setimbang yang optimal dengan konfigurasi teras 5×5 dan daya 20 MW, memiliki panjang operasi satu siklus lebih dari 40 hari. Perhitungan koefisien reaktivitas teras RRI dilakukan untuk bahan bakar baru U-9Mo-Al dengan kerapatan bervariasi. Perhitungan dilakukan dengan paket program WIMSD-5B dan BATAN-FUEL. Hasil pehitungan digunakan untuk melengkapi data desain konseptual teras yang menunjukkan bahwa teras setimbang reaktor RRI dengan konfigurasi 5×5, tingkat muat 235U sebesar 450 g, 550 g dan 700 g memiliki nilai koefisien reaktivitas temperatur bahan bakar, temperatur moderator, densitas moderator dan void semuanya negatif dan nilainya sangat bervariasi. Hal ini sudah memenuhi kriteria keselamatan desain konseptual teras RRI. Kata kunci: desain konseptual, bahan bakar uranium-molibdenum, koefisien reaktivitas, WIMS, BATANFUEL   The multipurpose of research reactor utilization make many countries build the new reserach reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research reactor in Indonesia right now is already 25 year old. Therefor, it is needed to design a new research reactor as a alternative called it innovative research reactor (IRR and then as

  9. RA reactor operation and maintenance in 1992, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktoro RA u 1992. Godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Majstorovic, D; Tanaskovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1992-12-01

    During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems. [Serbo-Croat] U toku 1992 godine poslovi u okviru projekta 'Istrazivacki nuklearni reaktor RA' obavljani su u skladu sa programom i planom rada. Osnovne aktivnosti na kojima je radjeno odnosile su se na revitalizaciju reaktora RA, kao i na odrzavanje opreme. U ovom periodu reaktor nije bio u pogonu. Svo osoblje je bilo angazovano na poslovima rekonstrukcije i modernizacije postojecih i dogradnje novih reaktorskih sistema, na odrzavanju opreme a deo tehnickog osoblja je bio obucavan za vrsenje odgovarajucih poslova u pogonu i odrzavanju opreme.

  10. Pengolahan Limbah Rumah Makan dengan Proses Biofilter Aerobik

    Directory of Open Access Journals (Sweden)

    Laily Zoraya Zahra

    2015-03-01

    Full Text Available Berkembangnya rumah makan/restoran yang semakin pesat dapat dipastikan akan turut menambah buangan air limbah domestik dengan kadar organik yang tinggi dalam jumlah yang tidak sedikit yang dibuang ke badan air.Tingginya kadar organik dalam limbah domestik rumah makan akan menyebabkan aroma yang tidak sedap jika tidak ada pengolahan terlebih dahulu, maka pengolahan yang dapat digunakan untuk mengolah air limbah rumah makan tersebut adalah proses biofilter aerobik. Penelitian dilakukan dengan proses biofilter aerobik dengan aliran downflow dan menggunakan sistem intermitten.Variabel dalam penelitian ini adalah media biofilter berupa kerikil dan batu alam serta Hydraulic Retention Time (HRT 8 jam. Parameter pencemar yang diukur efisiensinya adalah BOD, COD, dan TSS. Besarnya penyisihan parameter BOD, COD dan TSS dengan menggunakan biofilter aerob berturut-turut mencapai 94,83%, 92,95%, dan 95%. Reaktor paling baik dalam mengolah air limbah adalah reaktor biofilter dengan media kerikil pada HRT 8 jam.

  11. FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE

    Directory of Open Access Journals (Sweden)

    Entin Hartini

    2016-06-01

    BEJANA TEKAN REAKTOR: 2D DENGAN BEBAN INTERNAL PRESSURE. Bejana tekan reaktor (RPV merupakan pressure boundary dalam reaktor tipe PWR yang berfungsi untuk mengungkung material radioaktif  yang dihasilkan pada proses reaksi berantai. Maka dari itu integritas bejana tekan reaktor harus senantiasa terjamin baik reaktor dalam keadaan operasi normal, maupun kecelakaan. Dalam melakukan analisis integritas RPV, khususnya yang berkaitan dengan pecahnya bejana tekan reaktor akibat adanya retak dilakukan analisis secara fracture mechanics. Adanya ketidakpastian input seperti sifat mekanik bahan, lingkungan fisik, dan input pada data, maka dalam melakukan analisis keandalan tidak hanya dilakukan secara deterministik saja. Tujuan dari penelitian ini adalah melakukan analisis ketidakpastian input pada perhitungan fracture mechanik pada evaluasi keandalan bejana tekan reaktor PWR. Pendekatan untuk karakter random dari kuantitas input menggunakan  teori probabilistik. Analisis fracture mechanics dilakukan berdasarkan metode elemen hingga (FEM menggunakan perangkat lunak MSC MARC. Analisis ketidakpastian input dilakukan berdasarkan probability density function dengan Latin Hypercube Sampling (LHS menggunakan python script. Output dari MSC MARC adalah nilai J-integral untuk mendapatkan nilai stress intensity factor pada evaluasi keandalan bejana tekan reactor 2D. Dari hasil perhitungan dapat disimpulkan bahwa SIF probabilistik lebih dulu mencapai nilai batas fracture tougness  dibanding  SIF deterministik. SIF yang dihasilkan dengan metode probabilistik adalah 105,240 MPa m0,5. Sedangkan SIF metode deterministik adalah 100,876 MPa m0,5. Kata kunci: Analisis ketidakpastian, fracture mechanics, LHS, FEM, bejana tekan reaktor

  12. The reactor accident in the Three Mile Island-2 reactor plant. (Harrisburg, USA)

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The presentation of the accident development is based on the bulletin of the Atomic Industrial Forum (AIF) from the 6th of April, 1979. In addition, there are some short pieces of information, No. 14 and 15 of the association for reactor security as well as written and verbal information of the firms Brown, Boveri and Co/Babcock-Brown Boveri Reaktor GmbH (BBC/BBR). (orig.) [de

  13. Annual report of the group for maintenance of electrical equipment; Godisnji izvestaj elektro grupe

    Energy Technology Data Exchange (ETDEWEB)

    Rajic, M [Odelenje odrzavanja, Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1965-12-15

    This report includes detailed description of repairs and revisions of the electrical equipment of the RA reactor which were done according to the annual plan during the periods when reactor was not operated. Unplanned repairs are part of this report as well. [Serbo-Croat] Ovaj izvestaj sadrzi detaljan opis remontnih i revizionih radova na elektroopremi reaktora RA izvrsenih prema godisnjem planu u periodima kada reaktor nije radio. Vanplanski poslovi ukljucujuci popravke elektroopreme su takodje deo ovog izvestaja.

  14. Quality management in an international nuclear power plant project

    International Nuclear Information System (INIS)

    Brion, J.; Crustin, J.

    1975-01-01

    SNR (Schneller Natriumgekuehlter Reaktor) is the fast reactor power plant being erected at Kalkar, Federal German Republic. Quality management in this project is a contractual obligation. Quality management is subdivided into quality engineering, set of actions performed before manufacturing, and quality control, set of material controls performed during fabrication. The two successive phases of the quality management are presented. The difficulties and improvment possibilities are discussed [fr

  15. PENGOLAHAN LIMBAH CAIR INDUSTRI FARMASI FORMULASI DENGAN METODE ANAEROB-AEROB DAN ANAEROB-KOAGULASI

    OpenAIRE

    Farida Crisnaningtyas; Hanny Vistanty

    2016-01-01

    Studi ini membahas mengenai pengolahan limbah cair industri farmasi dalam skala laboratorium dengan menggunakan konsep anaerob-kimia-fisika dan anaerob-aerob. Proses anaerob dilakukan dengan menggunakan reaktor Upflow Anaerobic Sludge Bed reactor (UASBr) pada kisaran OLR (Organic Loading Rate) 0,5 – 2 kg COD/m3hari, yang didahului dengan proses aklimatisasi menggunakan substrat gula. Proses anaerob mampu memberikan efisiensi penurunan COD hingga 74%. Keluaran dari proses anaerob diolah lebih ...

  16. Pra Desain Pabrik Sorbitol dari Tepung Tapioka dengan Hidrogenasi Katalitik

    Directory of Open Access Journals (Sweden)

    Hellen Kartika Dewi

    2014-03-01

    Full Text Available Sorbitol yang dikenal juga sebagai glusitol, adalah suatu gula alkohol yang dimetabolisme lambat di dalam tubuh. Sorbitol banyak digunakan sebagai bahan baku untuk industri barang konsumsi dan makanan seperti pasta gigi, permen, kosmetika, farmasi, vitamin C, termasuk industri tekstil dan kulit. Pembuatan sorbitol dari bahan baku tepung tapioka. Pabrik sorbitol ini direncanakan akan didirikan di Propinsi Jawa Tengah tepatnya di Kabupaten Batang dengan kapasitas produksi 30.000 ton/tahun. Proses produksi Sorbitol menggunakan proses hidrogenasi katalitik. Pembuatan sorbitol dari bahan baku pati melalui dua tahap proses utama yaitu proses perubahan starch menjadi glukosa melalui hidrolisa double enzym. Enzim yang digunakan yaitu α-amylase dan glukoamylase. Proses hidrogenasi katalitik dilakukan dengan mereaksikan larutan dekstrose dan gas hidrogen bertekanan tinggi dengan menambahkan katalis nikel dalam reaktor (Reaktor Hidrogenasi. Gas hidrogen masuk dari bawah reaktor secara bubbling dan larutan dekstrose diumpankan dari atas reaktor sehingga kontak yang terjadi semakin baik. Sorbitol yang di hasilkan dalam pradesain pabrik sorbitol ini dengan konsentrasi 58,2%. Pendirian pabrik sorbitol memerlukan biaya investasi modal tetap (fixed capital sebesar Rp 168.801.192.952, modal kerja (working capital  Rp 29.788.445.815, investasi total Rp 198.589.638.767, Biaya produksi per tahun Rp 368.832.813.809 dan  hasil penjualan per tahun Rp 540.000.078.750. Dari analisa ekonomi didapatkan BEP sebesar 26,32%. ROI sesudah pajak 48,5 %, POT sesudah pajak 2,14 tahun. Dari segi teknis dan ekonomis, pabrik ini layak untuk didirikan.

  17. Inspection and repair of nuclear components

    International Nuclear Information System (INIS)

    Lahner, K.; Poetz, F.

    1993-01-01

    Despite careful design, manufacturing and operation, some of the important safety-relevant components show deterioration with time. Because of activation and contamination of these components, their inspection and repair has to be performed with manipulators. Some sophisticated manipulators are described, built by ABB Reaktor and used for inspection, maintenance and repair of PWR steam generators, fuel alignment pins, core baffle former bolts and reactor pressure vessel head penetrations. (Z.S.) 7 figs

  18. The training of the staff for work with radioactive materials and work on nuclear reactor in the Institute; Obuka kadrova za rukovanje radioizotopima i pogon nuklearnih reaktora u Institutu 'Boris Kidric' - Vinca

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M; Mladjenovic, O; Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1978-05-15

    A short informational review of the activities in the 'Boris Kidric' Institute on the training courses for the use of radioactive materials and for operating nuclear reactors including power reactors. The survey of the courses is given in the enclosures. (author) Kratak informativni pregled delatnosti u IBK na kursevima za obuku kadrova u rukovanju readioaktivnim materijalima i pogonu nuklearnih reaktora, ukljucujuci reaktore snage. pregled kurseva i materijala za njih dati su u prilozima. (author)

  19. Comparative studies in farther-reaching waste water cleaning in different reactor systems; Vergleichende Untersuchungen zur weitergehenden Abwasserreinigung in unterschiedlichen Reaktorsystemen

    Energy Technology Data Exchange (ETDEWEB)

    Dockhorn, T. [Technische Univ. Braunschweig (Germany). Inst. fuer Siedlungswasserwirtschaft

    1999-07-01

    Three semi-technical pilot plants (completely mixed reactor, cascade, SBR) were operated in parallel under equal starting conditions. The influence of the type of reactor on the processes COD elimination, nitrification, denitrification and biological P elimination under operating conditions was studied. (orig.) [German] Es wurden drei halbtechnische Versuchsanlagen (volldurchmischter Reaktor, Kaskade, SBR) unter gleichen Ausgangsbedingungen parallel betrieben. Hierbei wurde der Einfluss des Reaktortyps auf die Prozesse CSB-Elimination, Nitrifikation, Denitrifikation sowie biologische P-Elimination unter Betriebsbedingungen untersucht. (orig.)

  20. DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

    Directory of Open Access Journals (Sweden)

    Ralind Re Marla

    2015-03-01

    Full Text Available Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN untuk jenis Pebble Bed Modular Reactor (PBMR dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL. Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5 dan dari hasil analisis ini diharapkan dapat memenuhi syarat dalam mendukung program percepatan pembangunan kelistrikan batubara 10.000 MWe khususnya untuk proses smelter, yang tersebar merata di wilayah Indonesia. Hasil penelitian menunjukkan bahwa, faktor perlipatan efektif (k-eff Reaktor jenis PBMR daya 70 MWe mengalami kondisi kritis pada pengkayaan 5,626 % dengan nilai faktor perlipatan efektif 1,00031±0,00087 dan nilai koefisien reaktivitas suhu pada -10,0006 pcm/K. Dari hasil analisis daat disimpulkan bahwa reaktor jenis PBMR daya 70 MWe adalah aman.   ABSTRACT The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR type with 70 MWe capacity power in Beginning of Life (BOL has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff with power 70 MWe critical condition at enrichment 5,626 %is 1,00031±0,00087, based on enrichment result, a value of the temperature coefficient reactivity is - 10,0006 pcm/K. Based on the results of these studies, it can beconcluded that the PBMR 70 MWe design is theoritically safe.

  1. Activity of the RA Reactor Physics group in 1980 - Definition of the Operation conditions for future safe and economical RA reactor operation with 80% enriched fuel; Prilog Ia - Rad sluzbe za fiziku reaktora RA u 1980. godini - Definisanje pogonskih uslova za dalji siguran i ekonomican rad reaktora RA sa 80% obogacenim gorivom

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1980-12-15

    During 1980. the RA reactor was not in operation. That is why this period was devoted to definition of operating conditions for further reactor operation with 80% enriched fuel. The fuel elements which were in the core at the moment of shutdown in March 1979will not be used again (388 80% enriched fuel elements, and 511 2% enriched fuel elements). The reactor will be operated only with 80% enriched fuel, staring with initiat core configuration with 440 elements on the borders gradually changing to equi;librium core with 720 fuel elements. The analyses were concerned with safety issues of future operation. [Serbo-Croat] Reaktor RA tokom 1980 godine nije radio. Zbog toga je ovaj period iskoriscen za intenzivan rad na definisanju pogonskih uslova za dalji rad reaktora sa 80% obogacenim gorivom. Gorivo sa kojim je reaktor prestao da radi marta 1979. godine (388 gorivnih elemenata 80% obogacenoh i 511 elementa 2% obogaceno) nece biti vraceno u jezgro, vec ce reaktor raditi samo sa 80% obogacenim gorivom, pocev od konfiguracije sa 440 elemenata na periferiji do ravnotezne konfiguracije sa 720 elemenata. Sve nalize su se bavile sigurnosnim aspektima rada reaktora sa visokoobogacenim gorivom u jezgru.

  2. FACTORS INFLUENCING HUMAN RELIABILITY OF HIGH TEMPERATURE GAS COOLED REACTOR OPERATION

    Directory of Open Access Journals (Sweden)

    Sigit Santoso

    2016-10-01

    ABSTRAK Peran dan tindakan operator pada reaktor berpendingin gas akan berbeda dengan peran operator pada operasi tipe reaktor lain. Analisis unjuk kerja operator dan faktor yang berpengaruh dapat dilakukan secara komprehensif melalui analisis keandalan manusia(HRA. Melalui HRA dampak dari kesalahan manusia pada sistem maupun cara untuk mengurangi dampak dan frekuensi kesalahan dapat diketahui. Makalah membahas faktor yang berpengaruh pada tindakan operator, yaitu pada kejadian kecelakaan pendingin reaktor gas bersuhu tinggi-HTGR. Analisis untuk kualifikasi faktor pembentuk kinerja(PSF dilakukan berdasarkan kurva keandalan fungsi waktu, dan metode keandalan manusia yang dikembangkan berdasar pada aspek kognitif yaitu Cognitive Reliability and Error Analysis Method (CREAM. Hasil analisis berdasar kurva keandalan fungsi waktu menunjukkan komponen waktu berkontribusi positif pada peningkatan keandalan operator (PSF<1 pada kondisi semua fitur keselamatan berfungsi sesuai rancangan. Sedangkan pada metoda analisis dengan pendekatan kognitif CREAM diketahui selain faktor ketersediaan waktu, faktor pelatihan dan rancangan HMI juga berkontribusi meningkatkan keandalan operator. Faktor pembentuk kinerja keseluruhan diketahui sebesar 0,25 dengan faktor kontribusi positif dominan atau berpengaruh pada penurunan kesalahan manusia adalah ketersediaan waktu (PSF=0,01, dan faktor kontribusi negatif dominan adalah prosedur dan siklus kerja (PSF=5. Nilai PSF tersebut sebagai faktor pengali dalam perhitungan probabilitas kesalahan manusia. Analisis faktor pembentuk kinerja perlu dikembangkan pada skenario kejadian lain untuk selanjutnya digunakan untuk perhitungan dan analisis keandalan manusia yang komprehensif dan perancangan sistem interaksi manusia mesin di ruang kendali. Kata kunci: PSF, HTGR, operator, ruang kendali, keandalan manusia

  3. Penurunan Logam Timbal (Pb pada Limbah Cair TPA Piyungan Yogyakarta dengan Constructed Wetlands Menggunakan Tumbuhan Eceng Gondok (Eichornia Crassipes

    Directory of Open Access Journals (Sweden)

    Eko Siswoyo

    2015-10-01

    Full Text Available Salah satu permasalahan lingkungan yang ditimbulkan dari adanya lindi di TPA Piyungan yaitu pencemaran pada badan air, sungai dan air tanah. Untuk mengatasi permasalahan ini salah satunya dengan sistem Constructed Wetlands dengan menggunakan tumbuhah eceng gondok. Tujuan dari penelitian ini adalah untuk mengetahui tingkat penurunan konsentrasi Timbal (Pb yang terdapat dalam limbah cair TPA Piyungan dengan Constructed Wetlands menggunakan tumbuhan eceng gondok dan untuk mengetahui seberapa besar kapasitas serapan tumbuhan eceng gondok terhadap kandungan Timbal (Pb dalam limbah cair TPA Piyungan.Dalam penelitian ini digunakan reaktor yang terbuat dari kayu yang dilapisi plastik dengan ukuran 0,5 m x 1,0 m. Setiap reaktor diberi media tanah 5 cm, dan diberi tumbuhan sebanyak 14 buah. Reaktor tersebut diberi perlakuan dengan konsentrasi limbah yang bervariasi (100%, 75%, 50%, 25%, dan 0%, dan waktu pengambilan sampel (0, 3, 6, 9, 12 hari. Dengan menggunakan metode SSA (Spektrofotometri Serapan Atom.Berdasarkan pengujian diperoleh bahwa penurunan logam Pb pada limbah cair TPA Piyungan hari ke- 12, yaitu sebesar 0.0501mg/L pada konsentrasi 100%, 0.0295mg/L pada konsentrasi 75%, 0.0267mg/L pada konsentrasi 50% dan 0.0041 mg/L pada konsentrasi 25%.

  4. Kajian Pemilihan Sumber Mikroorganisme Solid Phase Microbial Fuel Cell (SMFC Berdasarkan Jenis dan Volume Sampah, Power Density dan Efisiensi Penurunan COD

    Directory of Open Access Journals (Sweden)

    Ganjar Samudro

    2017-06-01

    Full Text Available Mikroorganisme merupakan salah satu komponen penting dalam proses Solid Phase Microbial Fuel Cell (SMFC untuk degradasi bahan organik dan transfer elektron. Pemilihan sumber mikroorganisme menjadi metode yang paling sederhana untuk dikaji sebagai informasi awal ketersediaan dan identifikasi jenis mikroorganisme yang mendukung proses SMFC. Tujuan kajian ini adalah untuk memilih sumber mikroorganisme tanah, septic tank dan sedimen sungai yang tepat digunakan dalam proses SMFC berdasarkan jenis dan volume sampah, power density, dan efisiensi penurunan COD. Kajian ini didasarkan pada hasil penelitian menggunakan reaktor SMFC tipe single chamber microbial fuel cell dengan variabel jenis dan volume sampah , serta sumber mikroorganisme. Metode perbandingan secara kuantitatif dilakukan berdasarkan kecenderungan nilai power density dan efisiensi penurunan COD tertinggi di antara jenis dan volume sampah kantin, dedaunan dan komposit kantin-dedaunan. Hasil yang didapatkan adalah sumber mikroorganisme tanah dan sedimen sungai tepat digunakan untuk volume sampah 1/3 dan 2/3 dari volume reaktor, sedangkan sumber mikroorganisme septic tank tepat digunakan untuk volume sampah 1/3 dan 1/2 dari volume reaktor. Sumber mikroorganisme dari septic tank menunjukkan kinerja power density dan efisiensi penurunan COD yang lebih rendah dibandingkan sumber mikroorganisme tanah dan sedimen sungai.

  5. Wpływ czasu napowietrzania na pracę reaktora SBR i SBBR

    Directory of Open Access Journals (Sweden)

    Małgorzata Makowska

    2016-06-01

    Full Text Available W pracy przedstawiono efekty oczyszczania małych ilości ścieków w bioreaktorach porcjowych. Przebadano dwa równolegle pracujące reaktory, z czego jeden był klasycznym systemem SBR, a drugi to reaktor SBBR ze złożem ruchomym. Obydwa ciągi technologiczne oczyszczały taką samą ilość ścieków bytowych wstępnie podczyszczonych w piaskowniku. Trzy kolejne serie badań różniły się długością napowietrzania. Porównano pracę reaktorów w kolejnych seriach oraz analizowano wyniki uzyskane w obu systemach. Największą skuteczność usuwania związków węgla organicznego (jako ChZT i azotu (jako N-NH4 uzyskano w reaktorze SBBR w trzeciej serii badań (średnio odpowiednio 90 i 62%. Reaktor SBBR pracował bardziej stabilnie oraz usuwał na drodze biologicznej fosfor z największą średnią skutecznością 83% w serii trzeciej. Ścieki oczyszczone w tym reaktorze charakteryzowały się mniejszym stężeniem zawiesiny, co świadczy o skuteczniejszej pracy reaktora z wypełnieniem w fazie sedymentacji.

  6. ANALISA KARAKTERISTIK MINYAK PLASTIK HASIL DUA KALI PROSES PIROLISIS

    Directory of Open Access Journals (Sweden)

    Untung Surya Dharma

    2015-06-01

    Full Text Available Limbah plastik dapat dimanfaatkan sebagai bahan baku minyak plastik dengan menggunakan proses pirolisis. Minyak plastik yang dihasilkan dapat dimanfaatkan sebagai zat aditif atau campuran bahan bakar pada mesin. Pada Penelitian ini, proses pembuatan minyak plastik menggunakan dua kali proses pirolisis. Suhu reaktor pada proses pirolisis yang pertama dan kedua berbeda berturut-turut yaitu 200 oC dan 150 oC. Dari hasil penelitian ini diketahui bahwa pada proses pirolisis pertama dengan suhu reaktor 200 oC, dari 25 kg bahan baku menghasilkan 15,5 liter minyak plastik dalam waktu 80 jam. Sedangkan pada proses pirolisis kedua dengan suhu reaktor 150 oC, dari 15 liter minyak plastik dari hasil proses pirolisis pertama menghasilkan 11,6 liter minyak plastik dalam waktu 3,33 Jam. Adapun karakter minyak plastik yang dihasilkan adalah massa jenis 771,4 kg/m3, Viskositas 0,501 m2/s dan Nilai kalor 10518 kJ/kg

  7. ANALISIS SKENARIO KEGAGALAN SISTEM UNTUK MENENTUKAN PROBABILITAS KECELAKAAN PARAH AP1000

    Directory of Open Access Journals (Sweden)

    D.T. Sony Tjahyani

    2014-03-01

    Full Text Available Kejadian Fukushima telah menunjukkan bahwa kecelakaan parah dapat terjadi, maka dari itu sangatlah penting untuk menganalisis tingkat keselamatan pada reaktor daya. Berdasarkan rekomendasi expert mission IAEA setelah kejadian Fukushima, perlu dilakukan upaya untuk meminimalisasi terjadinya kecelakaan parah yaitu dengan melakukan proses pendinginan yang maksimal. Dalam konsep keselamatan fasilitas nuklir, khususnya reaktor daya telah diterapkan konsep keselamatan berlapis (Defence in Depth, DiD. Konsep keselamatan tersebut terdiri atas 5 level pertahanan yang bertujuan mencegah dan mengurangi lepasan produk fisi ke masyarakat dan lingkungan pada saat reaktor daya mengalami kecelakaan. Dalam reaktor telah didesain sistem atau tindakan yang mempunyai fungsi untuk mengatasi setiap level tersebut. Tujuan dari analisis ini adalah menentukan probabilitas kecelakaan parah dengan melakukan skenario kegagalan sistem dalam proses pendinginan di reaktor. Sebagai obyek analisis adalah reaktor daya AP1000, karena jenis reaktor ini sedang banyak dibangun saat ini. Skenario dilakukan dengan mengasumsikan beberapa kombinasi kegagalan sistem yang termasuk dalam DiD level 2 dan 3. Kegagalan sistem kemudian dianalisis dengan menggunakan analisis pohon kegagalan berdasarkan perangkat lunak SAPHIRE ver. 6.76. Dari analisis didapatkan probabilitas gagal dari kelompok sistem DiD level 2 dan 3 pada AP1000 masih di bawah batas kriteria dari IAEA yaitu lebih kecil dari 10-2, serta probabilitas kecelakaan parah didapatkan sebesar 6,17 x 10-10. Berdasarkan analisis ini disimpulkan bahwa AP1000 mempunyai tingkat keselamatan yang cukup tinggi, karena melalui skenario kegagalan sistem didapatkan probabilitas kecelakaan parah yang sangat kecil.   ABSTRACT Fukushima accident has shown that severe accident could be occurred, therefore it is important to analyze safety level of nuclear power plants. Based on the recommendations of IAEA expert mission after the Fukushima accident

  8. Lessons learned from full-scale vibration tests on nuclear power plant auxiliary structure in Switzerland

    International Nuclear Information System (INIS)

    Berger, E.; Tinic, S.

    1988-01-01

    The Beznau Nuclear Power Plant is located in northern Switzerland. The plant is owned and operated by the Nordostschweizerische Kraftwerke AG (NOK) in Baden, Switzerland. It is a twin unit plant (2 x 350 MWe) which was designed in the early 1960's and placed into commercial operation between 1969 and 1971. In connection with a major backfit project, which will improve the safety of the plant against external events, the free-standing boric water tanks had to be relocated and were replaced by two boric water tanks in a new building (the so called BOTA-building). It enabled to plan and perform full scale vibration tests.The scope of experimental investigation was to determine the eigenfrequencies and damping values for fundamental soil-structure interaction. The vibration tests allowed identification of the important modes of the soil-structure system in the range 3 to 15 Hz. The excitation was strung enough to generate accelerations in the structure comparable to those of a small earthquake. From the comparisons of computed and measured results it is concluded that the rocking frequency can be reasonably well predicted by either Finite Element or Lumped Parameter models with springs simulating the soil-foundation stiffness, provided in the case of the latter the embedment is taken into account. The prediction of the amplitude of structural response appears to be more difficult, as shown by the differences in the mode shapes. In the frequency range 8 to 10 Hz the agreement between computed and test results was less satisfactory. The actual structural behaviour turned out to be more complex than expected and needs further investigation with the aid of more refined models for the soil-structure system

  9. Lessons learned from full-scale vibration tests on nuclear power plant auxiliary structure in Switzerland

    Energy Technology Data Exchange (ETDEWEB)

    Berger, E [Basler and Hofmann AG, Consulting Engineers, Zurich (Switzerland); Tinic, S [Nordostschweizerische Kraftwerke AG, Baden (Switzerland)

    1988-07-01

    The Beznau Nuclear Power Plant is located in northern Switzerland. The plant is owned and operated by the Nordostschweizerische Kraftwerke AG (NOK) in Baden, Switzerland. It is a twin unit plant (2 x 350 MWe) which was designed in the early 1960's and placed into commercial operation between 1969 and 1971. In connection with a major backfit project, which will improve the safety of the plant against external events, the free-standing boric water tanks had to be relocated and were replaced by two boric water tanks in a new building (the so called BOTA-building). It enabled to plan and perform full scale vibration tests.The scope of experimental investigation was to determine the eigenfrequencies and damping values for fundamental soil-structure interaction. The vibration tests allowed identification of the important modes of the soil-structure system in the range 3 to 15 Hz. The excitation was strung enough to generate accelerations in the structure comparable to those of a small earthquake. From the comparisons of computed and measured results it is concluded that the rocking frequency can be reasonably well predicted by either Finite Element or Lumped Parameter models with springs simulating the soil-foundation stiffness, provided in the case of the latter the embedment is taken into account. The prediction of the amplitude of structural response appears to be more difficult, as shown by the differences in the mode shapes. In the frequency range 8 to 10 Hz the agreement between computed and test results was less satisfactory. The actual structural behaviour turned out to be more complex than expected and needs further investigation with the aid of more refined models for the soil-structure system.

  10. Karakterisasi Unjuk Kerja Diesel Engine Generator Set Sistem Dual Fuel Solar-Syngas Hasil Gasifikasi Briket Municipal Solid Waste (MSW Secara Langsung

    Directory of Open Access Journals (Sweden)

    Achmad Rizkal

    2017-01-01

    Full Text Available Sejalan dengan semakin banyaknya kebutuhan energi untuk dapat digunakan sebagai bahan bakar maka perlu adanya pengembangan gas biomassa sebagai bahan bakar alternatif pada motor pembakaran dalam maka akan dilakukan penelitian mengenai aplikasi sistem dual fuel gas hasil gasifikasi biomassa municipal solid waste (msw pada sistem downdraft dengan minyak solar pada motor diesel stasioner. Penelitian ini bertujuan untuk mengetahui seberapa besar solar yang tersibtitusi dengan adanya penambahan syngas yang disalurkan secara langsung. Penelitian ini dilakukan secara eksperimental dengan proses pemasukan aliran syngas yang dihasilkan downdraft municipal solid waste (MSW kedalam saluran udara mesin diesel generator set secara langsung menggunakan sistem mixer. Pengujian dilakukan dengan putaran konstan 2000 rpm dengan pembebanan bervariasi dari 200 watt sampai dengan 2000 watt dengan interval 200 watt. Bahwa produksi syngas dari reaktor gasifikasi ditambahkan sistem bypass untuk mengetahui kesesuaian antara reaktor gasifikasi dan mesin generatorset data ṁ syngas yang dibutuhkan mesin diesel, ṁ syngas yang di bypass untuk mendapatkan kesesuaian antara produksi syngas dan yang di bypass.  Data-data yang diukur dari penelitian ini menunjukkan bahwa besar nilai mass flowrate gas syngas yang dibutuhkan mesin diesel pada AFR reaktor gasifier 1,39 sebesar 0,0003748 kg/s. Mass flowrate gas syngas yang di bypass menunjukkan nilai 0 pada saat sistem dijalankan karena seluruh gas syngas masuk kedalam ruang bakar. AFR rata-rata sebesar 14,54 ,Nilai Spesifik fuel consumption (sfc mengalami peningkatan 68% dari kondisi standar single fuel , Nilai efesiensi thermal mengalami kenaikan sebesar 7% dari kondisi single fuel, Nilai daya rata-rata sebesar 2,28kW, Nilai torsi rata-rata sebesar 10,94 N.m. Solar yang tersibtitusi sebesar 48%. Nilai temperatur (coolant, mesin, oil, dan gas buang pada setiap pembebanan mengalami kenaikan.

  11. The development of quality assurance program in Reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Rosli Darmawan; Mohd Rizal Mamat; Mohamad Zaid Mohamad; Mohd Ridzuan Abdul Mutalib

    2007-01-01

    One of the trivial issues in the operation of Nuclear Reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing has always bring about general public fear on anything related to nuclear. IAEA has always emphasized on the assurance of nuclear safety for all nuclear installations and activities. According to the IAEA safety guides, all research reactors are required to implement quality assurance programs to ensure the conduct of operations are in accordance with the safety standards required. This paper discusses the activities carried out toward the establishment of Quality Assurance Program for Reaktor TRIGA PUSPATI (RTP). (Author)

  12. Design Of Pump Monitoring Of Primary Cooling System

    International Nuclear Information System (INIS)

    Indrakoesoema, Koes; Sujarwono

    2000-01-01

    Monitoring of 3 primary cooling pumps done visually by operator on the spot. The operator must be check oil in a sight glass, oil leakage during pump operation and water leakage. If reaktor power increase about more than 3 MW, the radiation exposure also increase in the primary cell and that's way the operator can not check the pumps. To continuing monitor all pump without delay, one system has been added I.e Closed Circuit Television (CCTV). This system using 3 video camera to monitor 3 pumps and connected to one receiver video monitor by coaxial cable located in Main Control Room. The sequence monitoring can be done by sequential switcher

  13. The treatment of uranyl nitrate from the AMOR process for VKTA Rossendorf

    International Nuclear Information System (INIS)

    Boessert, W.; Krempl, R.; Miller, J.W.

    2003-01-01

    The blending of uranyl nitrate solutions at VKTA and its subsequent treatment at BNFL-Sellafield is a significant step towards the safe and effective treatment of these enriched uranyl nitrate solutions. Overall the integration of the expertise of the international company BNFL/Westinghouse will lead to the achievement of a successful solution. This success has involved the integration of the project management and operational facilities of BNFL Sellafield with the local planning, design and manufacture capacities of Westinghouse Reaktor GmbH. (orig.)

  14. Simulation of atmosphere stratification in the HDR test facility with the CONTAIN code

    International Nuclear Information System (INIS)

    Skerlavaj, A.; Mavko, B.; Kljenak, I.

    2001-01-01

    The test E11.2 'Hydrogen distribution in loop flow geometry', which was performed in the Heissdampf Reaktor containment test facility in Germany, was simulated with the CONTAIN computer code. The predicted pressure history and thermal stratification are in relatively good agreement with the measurements. The compositional stratification within the containment was qualitatively well predicted, although the degree of the stratification in the dome area was slightly underestimated. The analysis of simulation results enabled a better understanding of the physical phenomena during the test.(author)

  15. Action plan during reactor shutdown in October 1965, Annex 5; Prilog br. 5 - Plan radova u toku stajanja reaktora u mesecu oktobru 1965. godine

    Energy Technology Data Exchange (ETDEWEB)

    Nikolic, M [Reaktor RA, Odelenje odrzavanja, Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1965-12-15

    The action plan of the division for reactor maintenance during reactor shutdown includes detailed list of tasks for mechanics, electronic and electrical equipment group during the reactor shutdown period in October 1965. It contains tasks for planned shutdown periods in September, August, July, May, April, March, and February 1965. [Serbo-Croat] Plan radova Odelenja odrzavanja reaktora RA za period stajanja reaktora u oktobru mesecu 1965. sadrzi detaljnu listu zadataka masinske grupe, elektro grupe i elektronske grupe. Ovaj prilog sadrzi i zadatke koji ce biti obavljeni tokom planiranih perioda kada je reaktor zaustavljen u septembru, avgustu, julu, junu, maju, aprilu, martu i februaru 1965.

  16. Nuclear Reactor RA Safety Report, Vol. 13, Causes of possible accidents

    International Nuclear Information System (INIS)

    1986-11-01

    This volume includes the analysis of possible accidents on the RA research reaktor. Any unwanted action causing decrease of integrity of any of the reactor safety barriers is considered to be a reactor accident. Safety barriers are: fuel element cladding, reactor vessel, biogical shield, and reactor building. Reactor accidents can be classified in four categories: (1) accidents caused by reactivity changes; (2) accidents caused by mis function of the cooling system; (3) accidents caused by errors in fuel management and auxiliary systems; (4) accidents caused by natural or other external disasters. The analysis of possible causes of reactor accidents includes the analysis of possible impacts on the reactor itself and the environment [sr

  17. Kinetics parameter measurements on RSG-GAS, a low-enriched fuel reactor

    International Nuclear Information System (INIS)

    Jujuratisbela, U; Arbie, B; Pinem, S.; Tukiran; Suparlina, L.; Singh, O.P.

    1995-01-01

    Kinetics parameter measurements, such as reactivity worths of control rods and fuel elements, beam tube void reactivity, power reactivity coefficient and xenon poisoning reactivity have been performed on different cores of Reaktor Serba Guna G.A. Siwabessy (RSG-GAS). In parallel, a programme was also initiated to measure the other kinetics parameters like effective delayed neutron life time, prompt neutron decay constant, validation of period reactivity relationship and zero power frequency response function. The paper provides the results of these measurements. (author)

  18. RA reactor operation and maintenance in 1994, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1994. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Mikic, N; Tanaskovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1994-12-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. The spent fuel elements used from the very beginning of reactor operation are stored in the existing pools. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988 and was fulfilled in 1990. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor. [Serbo-Croat] U proteklom periodu reaktor RA nije bio u pogonu zato sto je 30. jula Republicki komitet za zdravlje i socijalnu politiku republike Srbije, zabranio njegov rad zbog toga sto reaktor ne poseduje sistem za udesno hladjenje i ne poseduje odgovarajuce filtere u sistemu specijalne ventilacije. Zavrseni su radovi na izgradnji sistema za udesno hladjenje, rekonstrukciji postojeceg sistema specijalne ventilacije i rekonstrukciji sistema za napajanje elektricnom energijom. Zapoceti su radovi na modernizaciji, odnosno zameni instrumentacije reaktora ali njegova realizacija kasni

  19. Atomic Energy Research benchmark activity

    International Nuclear Information System (INIS)

    Makai, M.

    1998-01-01

    The test problems utilized in the validation and verification process of computer programs in Atomic Energie Research are collected into one bunch. This is the first step towards issuing a volume in which tests for VVER are collected, along with reference solutions and a number of solutions. The benchmarks do not include the ZR-6 experiments because they have been published along with a number of comparisons in the Final reports of TIC. The present collection focuses on operational and mathematical benchmarks which cover almost the entire range of reaktor calculation. (Author)

  20. Towards rational design of redox-stratified biofilms

    DEFF Research Database (Denmark)

    Lackner, Susanne

    Biologisk kvælstoffjernelse er en central proces indenfor avanceret spildevandsrensning. Denne afhandling beskriver anvendelsen af beluftede membran biofilm reaktorer (eng: MABRs) til fuld autotrof kvælstoffjernelse. Denne ret nye kvælstofomsætningsvej baseres på delvis omdannelse af ammonium til....... Som illustreret ved modelundersøgelserne, kan biofilm afrivning påvirke rektorfunktionen alvorligt. Den udførte forsøg for at vurdere om kemisk modificering af biofilmoverfladen kan øge dens modstandsdygtighed overfor afrivning og give forbedret biofilm tykkelseskontrol. Alt i alt kan MABR, hvor ilt...

  1. Possibilities for power reactor structural material and fuel testing in reactor RA; Mogucnosti reaktora RA za testiranje konstrukcionih materijala i goriva energetskih reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Lazarevic, Dj; Stefanovic, D; Cupac, S; Pesic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1978-05-15

    Nuclear reactor RA at Vinca has been designed as a high flux general purpose research reactor. Among other it was intended to play a role of material testing reactor. A scope of activities of Material Laboratory and Reactor RA Department of Boris Kidric Institute is presented in this report. Reactor RA capacity for reactor structural material and fuel irradiation is also described. The increase of RA reactor irradiation capacity is based on the improvement of VISA type fuel channel for fast neutron irradiations, as well as on the general neutron flux increase, due to introduction of highly enriched uranium fuel into reactor core and the advanced in-core fuel management. The irradiation capacities described allow for the reactor material and fuel testing to the considerable extent. Istrazivacki reaktor RA u Vinci je projektovan kao visokofluksni istrazivacki reaktor opste namene. Pored ostalog, on je namenjen i za testiranje reaktorskih konstrukcionih materijala i goriva. U radu je dat pregled aktivnosti Laboratorije za materijale IBK i reaktora RA na tom podrucju, kao i opis povecanih mogucnosti reaktora RA za ozracivanje reaktorskih materijala i goriva u cilju njihovog testiranja. Povecanje mogucnosti reaktora RA zasniva se na usavrsavanju specijalnog gorivnog kanala tipa VISA (za ozracivanje materijala brzim neutronima), kao i na opstem povecanju neutronskog fluksa na osnovu uvodjenja i nacina koriscenja visokoobogacenog uranskog goriva u reaktoru RA. Opisane mogucnosti reaktora RA dozvoljavaju u znatnoj meri ispitivanje konstrukcionih materijala i goriva energetskih reaktora.

  2. Efektivitas Pengolahan Air Limbah Batik dengan Cara Kimia dan Biologi

    Directory of Open Access Journals (Sweden)

    Istihanah Nurul Eskani

    2016-04-01

    Full Text Available Kebanyakan industri batik membuang air limbah ke lingkungan tanpa diolah terlebih dahulu dengan alasan keterbatasan tempat, dana dan penguasaan teknologi. Beberapa cara pengolahan air limbah telah dilakukan untuk mengatasi penurunan mutunlingkungan akibat pembuangan air limbah.Telah dilakukan penelitian proses pengolahan air limbah batik secara kimia, biologi aerob dan biologi anaerob. Proses kimia dilaksanakan dengan menambahkan koagulan tawas dan kapur ke dalam air limbah batik. Proses biologi aerob dijalankan dalam reaktor terbuka selama 5 hari, sedang proses biologi anaerob  dijalankan dalam reaktor tertutup selama 12 hari. Hasil proses kemudian diukur parameter warna, COD dan alkalinitasnya.Hasil penelitian pengolahan air limbah batik secara kimia dapat menurunkan parameter warna yang berasal dari zat warna Naphtol sebesar 83,15 %, COD sebesar 28,82% dan pH hasil proses 7. Proses biologi anaerob menurunkan parameter warna sebesar 94,95%, COD sebesar 59,89% dan pH hasil proses 5. Proses biologi aerob dapat menurunkan parameter warna yang berasal dari zat warna Naphtol sebesar 97,82%, COD sebesar 72, 88% dan pH hasil proses 6,5. Sehingga dapat disimpulkan bahwa pengolahan limbah cair batik secara biologi aerob lebih efektif daripada pengolahan secara biologi anaerob maupaun secara kimia. Kata kunci : pengolahan air limbah batik, proses kimia, proses biologi

  3. Data concerning operation and application of RA reactor in 1975, Annex 1; Prilog 1 - Podaci o radu i iskoriscenosti reaktora RA u 1975. godini

    Energy Technology Data Exchange (ETDEWEB)

    Stanic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-01-15

    RA reactor was operating according to the plan for 1975 adopted in December of the previous year. It was planned for reactor to be operated at nominal power first 10-20 days each month, three following days were reserved for different power levels according to the users' demand. Four fuel exchanges were planned and fulfilled with minor delay. Data concerning planned and real operation, as well as delays from the plan and shorter interruptions are presented in tables of this Annex. It is shown that all the delays and interruptions which amounted to 104 hours were compensated. [Serbo-Croat] Reaktor RA je radio prema planu rada za 1975. godinu, nacinjenom u decembru prethodne godine. Planirano je da reaktor radi neprekidno prvih 10-20 dana u mesecu na nominalnoj snazi, tri sledeca dana je rezervisano za rad na drugim snagama zavisno od potreba korisnuka. Planirane su i 4 izmene goriva. Podaci o planiranom i ostvarenom radu kao i odstupanjima od plana i kracim prekidima u radu reaktora dati su u tabelama ovog priloga. Vidi se da su sva odstupanja i prekidi u ukupnom trajanju od 104 sata u celini nadoknadjeni.

  4. ANALISA PENGASUTAN MOTOR INDUKSI 3 FASA 2500 KW SEBAGAI PENGGERAK FAN PADA BAG FILTER

    Directory of Open Access Journals (Sweden)

    Budi Yanto Husodo

    2017-11-01

    Full Text Available Salah satu persoalan yang timbul pada pengoperasian motor induksi adalah arus pengasutan yang tinggi yang nilainya bisa mencapai sepuluh kali arus nominal. Arus pengasutan yang besar ini mengakibatkan penurunan tegangan sesaat (sag pada sistem jaringan. Selain itu juga menyebabkan tingginya pemakaian daya hingga sebesar 1,5 - 2,5 kali daya nominal yang berakibat pada tingginya energi pemakaian pada saat pengasutan. Metode pengasutan diperlukan untuk mengurangi arus pengasutan dan pemakaian energi yang besar tersebut. Pada makalah ini dilakukan perbandingan tiga metode pengasutan motor induksi yaitu berupa autotrafo, reaktor dan rangkaian star-delta. Dengan menggunakan software ETAP, pengujian dilakukan pada motor induksi 6 kV, 279 A, 2500 kW, dan faktor daya sebesar 0,879, sebagai penggerak fan pada bag filter. Hasil simulasi menunjukkan bahwa autotrafo memberikan penurunan arus pengasutan yang paling besar yaitu 73,82% dari arus pengasutan motor tanpa bantuan alat pengasutan. Sedangkan konsumsi energi yang paling kecil didapatkan dengan mengunakan pengasutan reaktor, di mana energi pemakaian berkurang dari 31,102 kWh tanpa pengasutan menjadi 17,676 kWh atau setara dengan 43,17%.

  5. RA reactor operation and maintenance in 1989, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1989. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Zivotic, Z; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1989-12-15

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in July 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The following major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the power supply system. Project concerned with renewal of RA reactor complete instrumentation was started at the end of 1988. Contract was signed between the IAEA and Soviet Atomenergoexport for supplying the new instrumentation for the RA reactor. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988. In 1989, device for water purification designed by the reactor staff started operation and spent fuel handling equipment is being mounted. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor. [Serbo-Croat] U proteklom periodu reaktor RA nije bio u pogonu zato sto je 30. jula 1984. Republicki komitet za zdravlje i socijalnu politiku republike Srbije, zabranio njegov rad zbog toga sto reaktor ne poseduje sistem za udesno hladjenje i ne poseduje odgovarajuce filtere u sistemu specijalne ventilacije. Zavrseni su radovi na izgradnji sistema za udesno hladjenje, rekonstrukciji postojeceg sistema specijalne ventilacije i rekonstrukciji sistema za napajanje elektricnom energijom. Krajem 1988, medjunarodna agencija za atomsku energiju potpisala je ugovor sa sovjetskom firmom Atomergexport za izradu novog sistema instrumentacije. Sa ciljem da se poveca i efikasnije koristi prostor za skladistenje ozracenog goriva, 1987. godine zapoceta je realizacija projekata preciscavanja vode u bazenima za odlezavanje

  6. ILK statement on the consequences of the Chernobyl accident. Taking stock after twenty years

    International Nuclear Information System (INIS)

    2006-01-01

    The Chernobyl reactor accident was the consequence of a reactor design which was not inherently safe, and of a lack of 'safety culture'. The RBMK-type reactor (a Russian graphite-moderated light water reactor design: reaktor bolshoi moshnosty kanalny=high-power channel reactor) had not been designed to a satisfactory safety level, and the operating staff were not informed on the weak spots in plant design. The combination of these factors caused the worst nuclear accident, completely destroying the reactor. The consequences may be seen as the product of two severe accidents superimposed upon each other: the explosion of the reactor, and core melt-down associated with an intense, persistent fire of the graphite moderator. The Statement contains analyses of these points: Release, Propagation and Deposition of Radioactive Materials; Protective Measures; Impact on the Environment and Agriculture; Assessment of Radiation Exposure; Health Impact; Psychological and Societal Impacts; Potential Residual Risks. (orig.)

  7. DESAIN KONSEP TANGKI PENAMPUNG BAHAN BAKAR PASSIVE COMPACT MOLTEN SALT REACTOR

    Directory of Open Access Journals (Sweden)

    A. Hadiwinata

    2015-04-01

    Full Text Available Passive Compact Molten Salt Reactor (PCMSR merupakan pengembangan dari reaktor MSR. Desain reaktor PCMSR membutuhkan tempat khusus penampung sementara bahan bakar pada saat terjadi insiden, misalnya kecelakaan yang menyebabkan peningkatan suhu bahan bakar. Tangki penampung bahan bakar tersusun dari 3 bagian yang saling terhubung yaitu bagian penampung cairan bahan bakar, cerobong (chimney, dan penukar kalor. Dalam penelitian ini, tangki dimodelkan secara lump dan dilakukan variasi daya awal reaktor dan ketinggian cerobong. Syarat batas model ditetapkan suhu bahan bakar maksimum 1400 °C, yang didasarkan pada titik didih larutan garam LiF-BeF2-ThF4-UF4. Analisis dilakukan dengan cara menghitung rugi tekanan total dan transfer kalor untuk variasi daya awal antara 1800-3000 MWth dan ketinggian cerobong antara 1-10 m. Hasil penelitian menunjukan semakin besar daya reaktor, maka tinggi tangki penampung bahan bakar dan tinggi alat penukar kalor yang dibutuhkan akan semakin besar, tejadi kenaikan suhu fluida pendingin dan suhu udara pendingin, dan menyebabkan kenaikan laju aliran masa fluida pendingin, sedangkan laju aliran masa udara menurun. Peningkatan ketinggian cerobong menyebabkan ketinggian tangki penampung bahan bakar dan ketinggian alat penukar kalor semakin menurun, penurunan suhu fluida pendingin, tetapi suhu udara meningkat, dan menyebabkan peningkatan laju aliran masa fluida pendingin, tetapi laju aliran masa udara akan semakin menurun. Kata kunci: PCMSR, cerobong, alat penukar kalor, variasi daya.   The Passsive Compact Molten Salat Reactor (PCMSR reactor is developed from MSR reactor. The PCMSR reactor design requires special place to temporarily storage for reactor fuel when incident occurs, such as when there is an accident which caused the temperature of the fuel increases. The tank consist of three interconnected parts, the reservoir liquid fuel, chimney, and the heat exchanger. In this research, the tank system is modeled based on

  8. Vermikompos Sampah Kebun dengan Menggunakan Cacing Tanah Eudrilus eugeneae dan Eisenia fetida

    Directory of Open Access Journals (Sweden)

    Etik Rahmawati

    2016-04-01

    Full Text Available Durasi yang panjang diperlukan dalam pengomposan konvensional sampah organik yang memerlukan waktu selama 2-3 bulan. Pengurangan waktu pengomposan dapat dilakukan dengan digunakannya cacing sebagai dekomposer. Penelitian ini bertujuan untuk menentukan tingkat degradasi sampah kebun menggunakan proses vermikomposting dan menentukan pengaruh jenis cacing Eudrilus eugeneae dan Eisenia fetida. Empat reaktor berukuran 8 L digunakan dalam penelitian ini. Percobaan dilakukan secara duplo selama 60 hari. Parameter yang dianalisis pada penelitian ini adalah ammonia nitrogen (NH3-N, nitrat nitrogen (NO3-N, Total Kjeldahl Nitrogen (TKN, dan C/N. Hasil penelitian menunjukkan bahwa tingkat degradasi sampah kebun dengan pengolahan vermikomposting yang dapat dicapai adalah 64,94-72,52%. Produksi kompos yang lebih tinggi dengan penggunaan Eisenia fetida.

  9. German Light-Water-Reactor Safety-Research Program

    International Nuclear Information System (INIS)

    Seipel, H.G.; Lummerzheim, D.; Rittig, D.

    1977-01-01

    The Light-Water-Reactor Safety-Research Program, which is part of the energy program of the Federal Republic of Germany, is presented in this article. The program, for which the Federal Minister of Research and Technology of the Federal Republic of Germany is responsible, is subdivided into the following four main problem areas, which in turn are subdivided into projects: (1) improvement of the operational safety and reliability of systems and components (projects: quality assurance, component safety); (2) analysis of the consequences of accidents (projects: emergency core cooling, containment, external impacts, pressure-vessel failure, core meltdown); (3) analysis of radiation exposure during operation, accident, and decommissioning (project: fission-product transport and radiation exposure); and (4) analysis of the risk created by the operation of nuclear power plants (project: risk and reliability). Various problems, which are included in the above-mentioned projects, are concurrently studied within the Heiss-Dampf Reaktor experiments

  10. NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE

    Directory of Open Access Journals (Sweden)

    Tukiran Surbakti

    2016-03-01

    Full Text Available Abstract NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE. Research of UMo fuel for research reactor has been developing  right now. The fuel of  research reactor used is uranium low enrichment with high density. For supporting the development of fuel, an assessment of mini fuel in the RSG-GAS core was performed. The mini fuel are U7Mo-Al and U6Zr-Al with densitis of 7.0gU/cc and 5.2 gU/cc, respectively. The size of both fuel are the same namely 630x70.75x1.30 mm were inserted to the 3 plates of dummy fuel. Before being irradiated in the core, a calculation for safety analysis  from neutronics and thermohydrolics aspects were required. However, in this paper will discuss safety analysis of the U7Mo-Al and U6Zr-Al mini fuels from neutronic point of view.  The calculation was done using WIMSD-5B and Batan-3DIFF code. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the safety margin. Power density of U7Mo-Al mini fuel bigger than U6Zr-Al fuel.   Key words: mini fuel, neutronics analysis, reactor core, safety analysis   Abstrak ANALISIS NEUTRONIK ELEMEN BAKAR UJI MINI DI TERAS RSG-GAS. Penelitian tentang bahan bakar UMo untuk reaktor riset terus berkembang saat ini. Bahan bakar reaktor riset yang digunakan adalah uranium pengkayaan rendah namun densitas tinggi.  Untuk mendukung pengembangan bahan bakar dilakukan uji elemen bakar mini di teras reakror RSG-GAS dengan tujuan menentukan jumlah siklus di dalam teras sehingga tercapai fraksi bakar maksimum. Bahan bakar yang diuji adalah U7Mo-Al dengan densitas 7,0 gU/cc dan U6Zr-Al densitas 5,2 gU/cc. Ukuran kedua bahan bakar uji tersebut adalah sama 630x70,75x1,30 mm dimasukkan masing masing kedalam 3 pelat dummy bahan bakar. Sebelum diiradiasi ke dalam teras reaktor maka perlu dilakukan perhitungan keselamatan baik secara neutronik maupun termohidrolik. Dalam makalah ini

  11. Independent CO{sub 2} loop for cooling the samples irradiated in vertical experimental channels of the RA reactor, Vol. I; Nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u vertikalnim eksperimentalnim kanalima reaktora RA, Album I

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M; Pavlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-06-15

    Project 'independent CO{sub 2} loop for cooling the samples irradiated in the vertical experimental channels of the RA reactor' is presented in two volumes: volume I - head of the low temperature coolant loop for reactor RA, and volume II - Outer low-temperature reactor coolant loop. Volume I includes: the design specifications for the head of the low-temperature coolant loop, technical description, thermal calculation, calculations of mechanical loads, antireactivity and activation of the components of the coolant loop head, engineering schemes and drawings, cost estimation data. [Serbo-Croat] Projekat 'Nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u vertikalnim eksperimentalnim kanalima reaktora RA', sastoji se od dva albuma: album I - Glava niskotemperaturno rashladne petlje za reaktor RA, album II - Spoljno kolo niskotemperaturne rashladne petlje za reaktora. Album I sadrzi projektni zadatak glave niskotemperaturne petlje, tehnicki opis, termicki proracun, proracun mehanickih naprezanja, antireaktivnosti i aktivacije kontrukcionih elemenata glave petlje, konstrukcione seme i crteze glave petlje, predracun.

  12. Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

    Science.gov (United States)

    Karim, Julia Abdul

    2008-05-01

    The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.

  13. Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

    International Nuclear Information System (INIS)

    Karim, Julia Abdul

    2008-01-01

    The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained

  14. RA reactor operation and maintenance in 1990 with comparative evaluation from 1986-1990, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1990. godini, uz uporedni pregled za period 1986-1990. godina

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Vasovic, B; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1990-12-15

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The new emergency cooling system and the reconstruction of the existing ventilation system were finished in 1989, the conditions for further reactor operation were fulfilled. In the meantime new licensing regulations adopted in 1988 were not demanding the mentioned conditions for reactors operated at power less than 10MW, RA reactor power being 6.5 MW. But the reactor could not be restarted due to planned renewal of the reactor instrumentation. It is planned to exchange the complete instrumentation by the end of 1991. Training program for the staff operating and maintaining the reactor components was prepared in 1985. Reconstruction, modification and construction of components demanded new documentation needed for further safe reactor operation. New version of RA reactor safety report was finished in 1986 according to the recommendations of IAEA and licensing regulations of Yugoslavia. In 1989, new documents were written covering regulations and instructions for reactor operation. The new reactor experimental loop was designed in 1986, and constructed and tested in 1990. All the reactor components were maintained by specific reactor services. Financing of the reactor remains a permanent problem. [Serbo-Croat] U proteklom periodu reaktor RA nije bio u pogonu zato sto je 30. jula 1984. godine Republicki komitet za zdravlje i socijalnu politiku republike Srbije, zabranio njegov rad zbog toga sto reaktor ne poseduje sistem za udesno hladjenje i ne poseduje odgovarajuce filtere u sistemu specijalne ventilacije. Radovi na izgradnji sistema za udesno hladjenje i rekonstrukciji postojeceg sistema specijalne ventilacije zavrseni su 1989. godine. Uslovi za nastavak rada reaktora

  15. Preliminary studi on neutronic aspect of a conceptual design of the Kartini reactor base ADS facility

    International Nuclear Information System (INIS)

    Tegas Sutondo

    2012-01-01

    A preliminary study on neutronic aspect of a conceptual design of ADS facility with the basis of Kartini Reaktor, has been performed. The study was intended to see the feasibility from neutronic point of view of Kartini reactor, to be used as a small scale of NPP’s waste transmutation experimental facility. A SRAC code was used as the basis of calculations. The results indicate that the presence of minor actinides (MA) will give a positive reactivity, which tends to increase with the increase of MA concentrations. Based on the defined criteria of subcriticality and by considering the core power distributions and the level of reactivity contribution of MA element, it is concluded that Kartini reactor is potential enough to be used as an ADS experimental facility, mainly for MA concentration between 30 to 50 % of the assumed mixture of C-MA matrix. (author)

  16. PENURUNAN COD, TSS DAN TOTAL FOSFAT PADA SEPTIC TANK LIMBAH MATARAM CITRA SEMBADA CATERING DENGAN MENGGUNAKAN WASTEWATER GARDEN (Degradation of COD, TSS and Total Phosphate in Septic Tank Wastewater of Mataram Citra Sembada Catering Using Wastewater

    Directory of Open Access Journals (Sweden)

    Dradjat Suhardjo

    2008-07-01

    Full Text Available ABSTRAK  Sumber limbah berasal dari septictank industri restauran (catering Citra Sembada Catering, termasuk dalam kategori limbah domestik. Limbah tersebut banyak mengandung komponen yang tidak diinginkan bila dibuang ke badan air. Konsentrasi limbah yang masih di atas baku mutu, di antaranya akan memunculkan masalah pencemaran. Reaktor Wastewater Garden yang menggunakan krikil (0,5Cm-1cm dan 6 jenis tanaman yaitu : melati air (Echinodoras paleafias, Cyperus (Cyperus, Futoi (Hippochaetes lymnenalis, Pisang air (Typhonodorum indleyanum, Pickerel rush (Pontedoria cordata, Cattail (Typha latifulia. Penelitian ini bertujuan untuk mengetahui tingkat efektivitas reaktor Wastewater Garden, apabila digunakan untuk menurunkan konsentrasi Chemical Oxygen Demand (COD, Total Suspended Solid (TSS dan Fosfat Total sebagai faktor pencemar pada limbah industri restauran (Citra Sembada Catering yang tertampung pada septictank. Penelitian dilakukan dengan menggunakan reaktor Wastewater Garden dengan sistem batch dan dimensi reaktor lm x 0.5m x lm. Zona air limbah 75 cm, dan zona substrat atau krikil 80 cm, akar tanaman ditanam sedalam l0-15 cm. Metode penelitian yang digunakan berdasarkan SNI, di mana COD mengacu pada SNI 06-6989.2-2004 metode refluks tertutup secara spektrofotometri, TSS mengacu pada SK SNI M-03-1990-F metode pengujian secara gravimetri dan Fosfat total mengacu pada SNI M-52-1990-03 metode asam askorbat dengan alat spektrofotometer. Penelitian ini dilakukan selama 12 hari di mana setiap 3 hari sampel diambil pada outlel kemudian dianalisis. Berdasarkan hasil penelitian ini, diperoleh bahwa penggunaan wastewater garden pada limbah cair Mataram Citra Sembada Catering dapat menurunkan COD dengan efektivitas optimum 40,81% pada hari ke-6, penurunan TSS 89,l2% pada efektifitas optirnum hari ke-12 dan penurunan fosfat total dengan efektivitas optimum pada hari ke-6 yaitu sebesar 99,73 %. Tanaman dapat hidup dengan subur.   ABSTRACT  Wastewater

  17. Experimental verification of reflector savings calculated by REDIR code using two-group method; Eksperimentalna provera dvogrupnog racunanja reflektorske ustede koriscenog u REDIR-u

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1968-12-15

    Radial buckling and reflector savings for heavy water reactor with 2% enriched uranium fuel were measured and calculated by the REDIR code. A comparison of the obtained values is presented in this paper dependent on the reactor lattice pitch and reflector thickness. Experimental results obtained for lattice pitch of 16 cm prove the validity of applying the REDIR code for power reactors. U radu je dato uporedjenje izmedju izmerenih i teorijski izracunatih vrednosti (prema programu REDIR) radijalnih baklinga i reflektorske ustede za teskovodni reaktorski sistem sa 2% obogacenim uranskim gorivom u zavisnosti od koraka resetke i debljine reflektora. Rezultati dobijeni eksperimentima pri koraku resetke od 16 cm potvdjuju ispravnost primene programa REDIR za energetske reaktore. (author)

  18. Pengaruh Jenis Bahan pada Proses Pirolisis Sampah Organik menjadi Bio-Oil sebagai Sumber Energi Terbarukan

    Directory of Open Access Journals (Sweden)

    M. Sigit Cahyono

    2013-06-01

    Full Text Available Sampah organik merupakan potensi sumber energi yang melimpah di Indonesia. Sampah organik berupa daun dan ranting kering bisa dikonversi menjadi bahan bakar berupa bio-oil melalui proses fast pirolisis. Tujuan dari penelitian ini adalah untuk mengetahui pengaruh jenis bahan terhadap rendemen dan nilai kalor bio-oil yang dihasilkan dari proses pirolisis sampah organik. Bahan baku berupa daun dan ranting kering campuran tanaman angsana, mahoni dan mangga dengan komposisi daun bervariasi 0%, 50%, dan 100%, dipotong-potong dengan ukuran maksimal 10 cm. Kemudian bahan baku tersebut dipanaskan di dalam reaktor pirolisis pada suhu 500 C selama 1 jam. Hasil penelitian menunjukkan bahwa nilai kalor tertinggi (5175,35 J/g dan rendemen tertinggi (24,5% didapatkan pada bio-oil yang dihasilkan dari pirolisis ranting 100%. Kata kunci: Sampah Organik, Bio-oil, Pirolisis, Rendemen, Nilai Kalor

  19. Komposit Nano TiO2 Dengan PCC, Zeolit atau Karbon Aktif Untuk Menurunkan Total Krom dan Zat Organik Pada Air Limbah Industri Penyamakan Kulit

    Directory of Open Access Journals (Sweden)

    Bumiarto Nugroho Jati

    2012-04-01

    Full Text Available Telah dilakukan penelitian untuk menurunkan total krom dan zat organik pada limbah industri penyamakan kulit dengan menggunakan nano TiO2 yang dikompositkan dengan adsorben karbon aktif, zeolit, dan precipitated calcium carbonate (PCC dalam suatu reaktor fotokatalitik yang disusun secara batch dan dilengkapi dengan 6 buah lampu UV dan magnetic stirrer. Penurunan kadar krom total diukur dengan menggunakan Atomic Absorption Spectro-photometer (AAS dan penurunan zat organik dianalisa dengan menggunakan titrasi permanganatometri. Hasil penelitian menunjukkan pengolahan terbaik untuk penurunan kadar krom total adalah dengan menggunakan komposit TiO2:PCC = 8:2 yang dapat menurunkan total krom hampir 100% pada menit ke-170 dengan konsentrasi awal 214,35 mg/L. Untuk penurunan kadar zat organik, pengolahan terbaik dengan menggunakan komposit TiO2:PCC = 9:1 yang dapat menurunkan kadar zat organik hingga 100% pada menit ke-180. 

  20. RA reactor Action plan for 1978 - Annex VI; Prilog VI - Plan rada reaktora RA za 1978. godinu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-12-15

    This annex include a detailed monthly action plan of the RA reactor for 1978. It was planned that the reactor will operate at nominal and lower power levels for 158 days. The plan is made with the aim to achieve same annual neutron flux as in 1977, which is somewhat higher then the annual production in the period from 1968-1975. [Serbo-Croat] Ovaj prilog sadrzi detaljan plan rada reaktora RA za 1978. godinu. Planirano je da reaktor radi na nominalnoj snazi of 6,5 MW i na manjim snagama ukupno 158 dana. Plan je napravljen tako da se ostvari godisnja proizvodnja neutronskog fluksa u obimu koji je relaizovan 1977. godine sto je nesto vise od godisnje proizvodnje u periodu 1968-1975.

  1. The Ellweiler uranium plant - a demolition and recycling project

    International Nuclear Information System (INIS)

    Mika, S.; Rohr, T.; Seehars, R.; Feser, A.

    1999-01-01

    The uranium plant at Ellweiler, district of Birkenfeld, was used for the production and storage of uranium concentrates. The owner of the Ellweiler uranium plant (UAE), Gewerkschaft Brunhilde GmbH, ceased processing uranium ore and recycling in 1989 and has been in liquidation since September 1991. The State of Rhineland-Palatinate, had safety measures adopted in a first step, getting the plant into a safe state by former plant personnel. The entire plant was demolished in a second step. The contract for demolishing the former uranium plant was awarded to ABB Reaktor as the general contractor in August 1996. Demolition work was carried out between April 1997 and May 1999. A total of approx. 7900 Mg of material was disposed of. At present, recultivation measures are being carried out. (orig.) [de

  2. Fuel element transfer cask modelling using MCNP technique

    International Nuclear Information System (INIS)

    Rosli Darmawan

    2009-01-01

    Full text: 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. (author)

  3. Fuel Element Transfer Cask Modelling Using MCNP Technique

    International Nuclear Information System (INIS)

    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.

  4. Outlook on to fuel cycle perspectives at WWER-440

    International Nuclear Information System (INIS)

    Stech, S.; Bajgl, J.

    2005-01-01

    Current internal fuel cycle in NPP Dukovany 4x440 MWe is shortly characterized with new types of fuel assemblies and advanced fuel cycles which have been introduced in the last years. The modernization activities accomplished until now might be extrapolated to the further period in fuel design - mechanic, thermal-hydraulic and neutronic respectively - with additional increase in fuel enrichments and burnups on the way to the 6-year cycle. Reaktor power up rating together with Unit thermal efficiency improvements could bring an increase in the electric output to the value nearly 500 MWe. The reasons are given for long-term cooperation with Fuel Supplier and Plant Designer in the area of fuel cycle as well as in Unit Design Basis. All innovations mentioned in the article including future fuel and fuel cycle changes might be a quite realistic perspective at the end of the first decade of the new century (Authors)

  5. Determination of bedrock depth by using microtremor array survey at the RDE Site Serpong

    International Nuclear Information System (INIS)

    Hadi Suntoko; Sriyana

    2016-01-01

    Based on the Strategic Spatial Planning (RTRW) of Tangerang Selatan 2011-2031, PUSPIPTEK Serpong has been appointed as high technology zone, therefore this area is selected as the site for Experimental Power Reactor (Reaktor Daya Eksperimental, RDE). In order to assure safety, site evaluation from several aspects have to be performed, one of which is seismological aspect. Refer to BAPETEN Chairman Regulation No. 8 year 2013, site evaluation should cover the assessment of subsurface geological condition of the site. The objective of this study is to acquire information regarding the subsurface geological condition particularly the bedrock depth of the RDE site area. Method used in this study is micro tremor array (MA). The result shows that RDE site has a bedrock type of Bojongmanik Formation at a depth of 391 meter from the surface. (author)

  6. Operation and maintenance of the RA reactor, RA Research reactor. Annual report 1976; Pogon i odrzavanje reaktora RA, Izvestaj o radu u 1976. godini

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-12-15

    During 1976 the Ra reactor was operating for about 30% shorter period than usual. The reason were extraordinary repair activities within regular and investment maintenance as well as repair of failures caused by neglected maintenance during previous 6 years. Delay was caused by unavailability of fuel (2% enriched fuel elements are spent) and the new 80% enriched fuel demanded experimental and theoretical analyses before being introduced into the core. Safety analyses concerned with using 80% enriched fuel both experimental and theoretical were successfully fulfilled. The December 1976 successful experimental campaign can be marked as end of the 17 years period of using 2% enriched fuel and start of the new period of using highly enriched fuel. This is significant not only for the reactor itself but for the users, because it would result in increase of neutron flux by 50% with the increase of costs by only 4%. Demand was submitted for obtaining the final license for transition operating regime with highly enriched fuel which would save at least 2 200 000 dinars. This will enable reactor operation in 1977 and later on, without interruption by 'critical' and other experiments related to new highly enriched fuel. A high number of repair and other urgent activities were fulfilled in order to enable safe operation. Some of these activities were done never before and some were neglected during past 6 years. The most important tasks were: purchase of Al tubes made of special alloy, fabrication and mounting of the fuel channel; overall investigation of reactor vessel leakage; repair of the heavy water pump; exchange of two vertical channels. basic equipment for construction of emergency cooling system was purchased. Hot cells are equipped for independent utilisation. [Serbo-Croat] Reaktor RA je u 1976. godini radio za oko 30% krace vreme od uobicajenog. Razlog su bill izuzetno veliki obim remontnih i drugih radova u okviru tekuceg i investicionog odrzavanja, kao i krupnih

  7. Activities of the Service for maintenance of the RA reactor electronic equipment in 1979 - Report - Annex IV; Prilog IV Rad sluzbe za odrzavanje elektronske opreme reaktora RA u 1979. godini - Izvestaj -

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1979-12-15

    Within the organizational structure of the RA reactor staff, the Service for instrumentation maintenance has the following tasks: maintenance of the existing electronic equipment; participating in experiments planning and preparation of electronic equipment; purchasing new equipment, spare parts and components; construction of new equipment for internal needs; implementation of new equipment. Basic instrumentation of the reactor facility includes: control and protection system, and dosimetry system. [Serbo-Croat] U okviru organizacione seme OOUR nuklearni reaktor RA sluzba za odrzavanje instrumentacije postrojenja ima sledece zadatke: odrzavanje postojece elektronske opreme; ucesce u planiranju eksperimenata i priprema elektronske opreme; nabavka nove opreme, rezervnih delova i komponenti; rad na izradi elektronske opreme i uredjaja za sopstvene potrebe i izrada projekata u domenu reaktorske opreme i ucesce u ugradnji nove opreme. Osnovna instrumentacija postrojenja obuhvata: sistem za upravljanje i zastitu, sistem za tehnolosku kontrolu i sistem za dozimetrijsku kontrolu.

  8. Fission gas release from the sintered UO{sub 2} fuel; Oslobadjanje fisionih gasova iz goriva od sinterovanog UO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Sigulinski, F; Stevanovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-11-15

    This paper shoes the phenomena which control fission gases release from the sintered UO{sub 2} dependent of the burnup rate: ejection, release, diffusion, increased fission gas accumulation causing structural changes in the fuel. release of fission gases from the fuel for power reactors was studied as well. The influence of factors as temperature, characteristics of fuel, burnup rate and burnup level was analyzed. Prikazani su mehanizmi koji kontrolisu izdvajanje fisionih gasova iz sinterovanog UO{sub 2} pri razlicitim brzinama izgaranja: izletanje, izbijanje, difuzija, povecano izdvajanje fisionih gasova koje prati strukturne promene u gorivu. Razmatrano je proucavanje izdvajanja fisionih gasova iz goriva za reaktore snage. Analiziran je uticaj faktora kao sto su temperatura, karakteristike goriva, brzina i stepen izgaranja (author)

  9. Definition of task (Phase II) B, 'Leak testing'; Definicija zadatka (II faza) B, 'Ispitivanje hermeticnosti'

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M; Nikolic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-01-15

    In order to ensure safe Ra reactor operation and fulfilling the conditions for performing irradiation experiments it was necessary to test and verify the leak tightness of the sample containers and thermocouples, as well as the irradiation capsule before it was placed in the VISA-2 channel. Leak testing of VISA-2 channel with capsules and thermocouples was done before and after it was built-in the reactor. [Serbo-Croat] U cilju bezbednosti i sigurnosti rada reaktora RA i ispunjavanja uslova ozracivanja izvrsena je provera hermeticnosti kenera sa uzorcima i termoparovima, kao i hermeticnost kapsule pre ugradjivanja kapsule u kanal VISA-2. Izvrseno je i ispitivanje hermeticnosti kanala VISA-2 sa kapsulama i termoparovima pre ugradjivanja u reaktor i posle montaze.

  10. Progress in the development of uranium silicide (U3Si2) fuel at BATAN

    International Nuclear Information System (INIS)

    Suripto, A.; Soentono, S.

    1995-01-01

    After successful fabrication of two full-size prototype fuel elements containing ∼3.0 gU/cm 3 in the form of U 3 Si 2 -Al dispersion now undergoing irradiation in the Reaktor Serba Guna G.A. Siwabessy (RSG-GAS) core since 1990, further development in U 3 Si 2 -A2 dispersion fuel element manufacturing has been pursued, whose progress in discussed in this paper, with a special attention on the use of much higher-loading aimed at obtaining a better understanding on the influence of higher-loading on fuel core and plate manufacturing and quality. At present, high-loading U 3 Si 2 -AI dispersion miniplates are being manufactured for preparing some mini-fuel elements to be test-irradiated in the new MTR in-pile loop of the RSG-GAS. (author)

  11. MASYARAKAT MADURA DAN RENCANA PEMBANGUNAN PLTN; PERSPEKTIF TEOLOGI

    Directory of Open Access Journals (Sweden)

    Abd. A'la

    2012-05-01

    Full Text Available Abd A’la   (Staf pengajar pada program Pascasarjana IAIN Sunan Ampel dan Asiten Direktur Bidang Akademik pada lembaga yang sama       Abstrak : Setiap pemanfaatan energi selalu memiliki resiko environmental. Resiko dalam dunia ilmu dapat ditolelir jika hal itu telah terkalkulasikan sehingga merupakan caculable risk. Sementara untuk energi nuklir, sampai saat ini, belum dapat dilakukan terutama terkait dengan teknologi pengamannya sehingga Fitjof Capra menyebutnya sebagai rasionalitas yang tidak bertanggung jawab. Masyarakat Madura merupakan komunitas Sunny dan dalam merespon setiap perkembangan selalu berjangkar pada prinsip-prinsip Sunny yang dicirikan dengan kehati-hatian dan moderasi. Terkait dengan rencana PLTN madura, masalah data tentang kebutuhan hal itu harus segera diperoleh untuk menjawab perlu tidaknya PLTN di Madura. Kekhwatiran bahwa kebutuan itu hanya merupakan rekayasa perusahaan reaktor nuklir patut di pertimbangkan karena memang sangat beralasan Kata Kunci : masyarakat madura, nuklir, PLTN

  12. ANALISA THERMOGRAVIMETRY PADA PIROLISIS LIMBAH PERTANIAN

    Directory of Open Access Journals (Sweden)

    Bagus Setiawan

    2016-01-01

    Full Text Available Penelitian ini bertujuan untuk menemukan karakterisasi degradasi termal dari limbah pertanian untuk dijadikan suatu bahan bakar padat alternatif. Penelitian diawali dengan tahap pengumpulan bahan yang dilanjutkan penyeragaman ukuran sampel uji hingga berukuran 20 mesh. Setelah itu masing-masing sampel dikeringkan hingga kadar air maksimal 12 %. Sebelum Sampel seberat 20 gram diuji pirolisis dengan menempatkan sampel dalam reaktor yang telah dialiri nitrogen dengan laju 100 ml/menit. Sampel diuji dengan kondisi heating rate 15 oC/menit, temperatur akhir 600 oC dan holding time 10 menit. Data yang didapat berupa penurunan massa dan perubahan temperatur dicatat dalam laptop dengan menggunakan software RS-Key, Ms Excel dan Adam.NET Utility. Dari penelitian yang telah dilakukan, maka dapat disimpulkan campuran serbuk gergaji dan jerami memiliki temperatur pirolisis paling rendah, sementara campuran sekam padi dan kulit singkong memiliki massa arang paling banyak.

  13. Anaerobic treatment of sulfate-containing wastewater from distilleries

    International Nuclear Information System (INIS)

    Stadlbauer, E.A.; Oey, L.N.; Weber, B.

    1994-01-01

    Bioprocess evaluation of a staged arrangement of a Pulse Driven Loop Reaktor (PDLR) and a Pulsed Anaerobic Filter (PAF) using highly polluted cherry slops as industrial wastewater shows a COD removal efficiency of 80-90% at loading rates of 8-4 kg COD/(M 3 .d). Contamination of cherry slops by sulfate (2 g/l) and copper (150-200 mg/l) reduces COD degradation to 40-50 percent. A pulsed anaerobic baffled reactor was envisaged as a corrective tool to improve mineralisation in the presence of sulfate-rich substrates by confining sulfate reducing bacteria to the first 4 chambers of the reactor. Phasing slightly improves COD degradation yield, but is not sufficient for stable process performance. Consequently, the use of lactic acid in stead of sulfuric acid in cherry-fermentation was suggested as a preventive method to avoid sulphide-induced digester failure. (orig.) [de

  14. Nuclear Reactor RA Safety Report, Vol. 1, Introduction; Izvestaj o sigurnosti nuklearnog reaktora RA, Knjiga 1, Uvod

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-11-01

    Based on the agreement between governments of Socialist Federal Republic of Yugoslavia and USSR of January 28 1956, a contract was signed about construction of RA research reactor in the Boris Kidric Institute of nuclear sciences. Building of the RA reactor started in 1956, and has reached criticality in 1959. Since then it has been in almost permanent operation, except for five longer shutdown periods: in 1963 because of heavy water and primary coolant system contamination with cobalt; in 1970 because of transporting the heavy water to France for isotopic regeneration; in 1979/1980 and 1983 because of aluminium oxyhydrate deposition on the fuel element cladding in reactor active region; in 1985/1986 because of ventilation system reconstruction and construction of emergency core cooling system. RA reactor is a heavy water cooled and heavy water moderated research reactor. Since the beginning of its operation, 2% enriched metal uranium fuel was used. From 1976, 80% enriched uranium oxide fuel elements were used partially in some core regions and since 1981 the complete core was filled with this highly enriched fuel. RA reactor was designed to operate under normal conditions at 6.5 MW power and at 10 MW power under forced regime. As a powerful neutron source the reactor was meant to be used for research in the field of reactor and neutron physics, solid state physics, radiation chemistry, biology and radioactive isotopes production. RA reactor was build by Yugoslav companies based on USSR basic design project. Main components of the systems were produced in USSR. [Serbo-Croat] Na osnovu Sporazuma izmedju vlada SFRJ i SSSR od 28. januara 1956. godine, sklopljen je ugovor o izgradnji istrazivackog reaktora RA u Institutu za nuklearne nauke Boris Kidric u Vinci. Reaktor RA je poceo da se gradi 1956. godine, a kriticnost je postigao 1959. godine. Od tada je takoreci neprekidno bio u radu, izuzev u pet slucajeva duzeg stajanja: 1963. godine - zbog kontaminacije teske vode

  15. RA reactor operation and maintenance, Annual report 1974; Pogon i odrzavanje reaktora RA - Izvestaj o radu u 1974. godini

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1974-12-15

    During 1974, RA reactor was operated at nominal power for 194 days and 13 days at lower power levels. The total production was 30711 MWh which is 2.4% higher than planned. Practically there were no discrepancies from the plan. The reactor was used for irradiation and experiments according to the demand of 437 users. This report contains detailed data about reactor power and experiments performed in 1974. Total number of afety shutdowns was 11, of which 8 were caused by power cuts, and 3 due to human error. Maximum individual personnel exposure dose was 50% of the maximum permissible dose. There were no accidents during this year. Decontamination of surfaces was less than during previous years. About 805 m{sup 2} of surfaces and 178 objects were decontaminated. It was concluded that the successful operation in 1974 has a special significance taking into account the financial problems. [Serbo-Croat] Reaktor RA je u 1974. godini radio na nominalnoj snazi 194 dana i 13 dana na manjim snagama. Ukupni rad iznosio je 30711 MWh odnosno 1,4% vise od planiranog. Prakticno nije bilo odstupanja od plana rada. Reaktor je koriscen za ozracivanja i eksperimente za 437 korisnika. Ovaj izvestaj sadrzi detaljne podatke o radu i eksperimentima koji su obavljani. U toku godine bilo je 11 sigurnosnih zaustavljanja, od cega 8 zbog elektricnog napona i 3 usled greske osoblja. Ukupna doza ozracivanja ljudstva bila je manja nego prethodnih godina. Maksimalna doza po oveku bila je 50% manja od maksimalno dozvoljene doze. Nije bilo ni jednog akcidenta. Dekontaminirano je znatno manje povrsina nego ranijih godina, i sakupljeno manje otpada nego prethodnih godina, dok tecnih efluenata nije bilo. Zakljuceno je da uspesan rad reaktora u 1974. godini ima poseban znacaj kada se imaju na umu problemi finansiranja reaktora.

  16. Report on the activity of the RA reactor operation for the period from July 1 1961 - Sept. 30 1961; Tromesecni izvestaj za period od 1.VII do 30.IX 1961. g

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V [Laboratorija za eksploataciju reaktora RA, Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-09-15

    During the reporting period the reactor was permanently ready for operation and responding to the demands of the experimenters. The reactor was operating for 408.5 hours at power levels from 50 - 5000 kW, or 1985 MWh in total, burnup of the first batch of fuel was 22.55%. Reactor core was made of 56 fuel channels. Activities related to construction of new and improvement of the existing equipment was continued in order to enable safe operation and successful utilization of the RA reactor. Exchange of the electronic tubes was continued in order to increase the stability of the reactor control and reactor protection systems. About 65% of tubes planned to be exchanged this year was done. Cooperation with the CEN Saclay, France related to construction of experimental loops VISA-1 and VISA-2 was continued as well as cooperation with Poland concerned with exchange of experts. The problem of lack of properly trained staff was nor solved. [Serbo-Croat] U izvestajnom periodu reaktor je uvek bio spreman za rad i odgovarao je zahtevima eksperimentatora. Reaktor je radio 408,5 casova na snagama od 50-5000 kW, odnosno ukupno 1985 MWh, pri cemu je ukupan utrosak prve sarze goriva bio 22,55%. Jezgro reaktora bilo je sacinjeno od 56 tehnoloskih kanala. Radi bezbednijeg pogona i sto uspesnije eksploatacije reaktora RA nastavljeno je sa radovima na realizaciji novih i poboljsanju postojecih uredjaja. U cilju povecanja stabilnosti elektronskih instrumenta u sistemu upravljanja i zastiti reaktora nastavljeno je sa zamenom elektronskih cevi, do sada je zamenjeno oko 65% radova predvidjenih za ovu godinu. Nastavljena je saradnja sa nuklearnim centrom u Saclay, Francuska na izradi petlje VISA-1 i VISA-2 i saradnja sa Poljskom u razmeni strucnjaka. Problem nedostatka kadrova nije resen.

  17. Commissioning Experience from the Agesta Nuclear Power Plant; Experience acquise lors des essais de mise en service de la centrale nucleaire d'Agesta; Opyt po vvedeniyu v ehkspluatatsiyu yadernoj ehnergeticheskoj ustanovki Agesta; Experiencia adquirida con la puesta en marcha de la central nucleoelectrica de Agesta

    Energy Technology Data Exchange (ETDEWEB)

    Rydell, N. [Aagesta Kraftvarmewerk, Farsta (Sweden)

    1963-10-15

    The Agesta Nuclear Power Plant is a pressurized heavy water reactor of the pressure vessel type, fuelled with natural uranium. It was commissioned with light water from December 1962 to May 1963. Observations of a more general interest were made during this commissioning essentially on the following topics; (a) cleanliness of primary circuit (b) valve operation (c) pressurization of the primary circuit (d) water leakage (e) refuelling machinery (f) containment testing. (author) [French] Il s'agit d'un reacteur a uranium naturel et a eau lourde pressurisee, du type a caisson sous pression. Les essais de mise en service ont ete faits avec de l'eau ordinaire, de decembre 1962 a mai 1963. La mise en service a permis de faire des observations d'interet general sur les sujets suivants: a) non-contamination du circuit primaire; b) fonctionnement des vannes; c) maintien sous pression du circuit primaire; d) fuites d'eau; e) appareils de chargement du combustible; f) essais d'isolement. (author) [Spanish] La central nucleoelectrica de Agesta posee un reactor de agua pesada del tipo de recipiente de presion, con combustible de uranio natural. Se mantuvo en funcionamiento con agua ligera entre diciembre de 1962 y mayo de 1963. Durante esta prueba, se efectuaron observaciones de interes mas general, relacionadas esencialmente con las siguientes cuestiones: a) limpieza del circuito primario; b) funcionamiento de las valvulas; c) presion del circuito primario; d) perdidas de agua; e) dosposiciones de reposicion del Combustible; f) ensayos de confinamiento. (author) [Russian] Yadernaya ehnergeticheskaya ustanovka Agesta predstavlyaet soboj tyazhelovodnyj reaktor pod davleniem, ispol'zuyushchij prirodnyj uran v kachestve topliva. Reaktor byl vveden v ehkspluatatsiyu na obychnoj vode v period s dekabrya 1962 goda po maj 1963 goda. Zamechaniya bolee obshchego kharaktera byli sdelany vo vremya ehkspluatatsii v osnovnom po sledukhshchim temam: a) chistota pervichnogo kontura; b

  18. ANALISA PENGARUH SUHU AWAL PELAT PANAS PADA PROSES QUENCHING CELAH SEMPIT REKTANGULAR

    Directory of Open Access Journals (Sweden)

    M. Hadi Kusuma

    2015-03-01

    Full Text Available Pemahaman terhadap manajemen termal apabila terjadi suatu kecelakaan parah reaktor nuklir seperti melelehnya bahan bakar dan teras reaktor, menjadi prioritas utama untuk menjaga integritas bejana tekan reaktor. Dengan demikian hasil lelehan bahan bakar dan teras reaktor (debris tidak keluar dari bejana tekan reaktor dan mengakibatkan dampak lain yang lebih besar ke lingkungan. Salah satu cara yang dilakukan untuk menjaga integritas bejana tekan reaktor adalah dengan melakukan pendinginan terhadap panas berlebih yang dihasilkan akibat dari kecelakaan tersebut. Untuk mempelajari dan mendapatkan pemahaman mengenai hal tersebut, maka dilakukan penelitian mengenai pengaruh suhu awal pelat panas dalam proses quenching (pendinginan secara tiba-tiba celah sempit rektangular. Penelitian difokuskan pada penentuan suhu rewetting dari pendinginan pelat panas dengan suhu awal pelat 220 0C, 400 0C, dan 600 0C dengan laju aliran air pendingin 0,2 liter/detik. Eksperimen dilakukan dengan menginjeksikan air pada laju aliran 0,2 liter/detik pada suhu air pendingin 85 0C ke dalam celah sempit rektangular. Data hasil pengukuran digunakan untuk mengetahui suhu rewetting yang terjadi pada pendinginan pelat panas tersebut. Tujuannya adalah untuk memahami pengaruh suhu awal pelat panas terhadap rewetting pada proses quenching di celah sempit rektangular. Hasil yang diperoleh menunjukkan bahwa titik rewetting pada pendinginan pelat panas 220 0C, 400 0C, dan 600 0C terjadi pada suhu rewetting yang berbeda-beda. Pada suhu awal pelat panas 220 0C, suhu rewetting terjadi pada 220 0C yaitu langsung ketika air dilewatkan melalui celah sempit rektangular. Pada suhu awal pelat panas 400 0C, suhu rewetting terjadi pada 379,51 0C. Dan pada suhu awal pelat panas 600 0C, suhu rewetting terjadi pada 426,63 0C. Perbedaan suhu awal pelat panas yang sangat signifikan menyebabkan terjadinya perubahan sifat fisik benda uji, berbedanya rejim pendidihan yang dialami oleh fluida yang

  19. INVESTIGATION ON THERMAL-FLOW CHARACTERISTICS OF HTGR CORE USING THERMIX-KONVEK MODULE AND VSOP'94 CODE

    Directory of Open Access Journals (Sweden)

    Sudarmono Sudarmono

    2015-03-01

      Kegagalan sistem pembuangan panas pada reaktor berpendingin air jenis PWR, Three Mile Islands dan reaktor BWR Fukushima Daiichi, menyebabkan masyarakat nuklir mulai memikirkan penggunaan reaktor pembangkit daya jenis temperatur tinggi berpendingin gas (HTGR. Bidang Fisika dan Teknologi Reaktor di Pusat Teknologi Reaktor dan Keselamatan Nuklir (PTRKN mempunyai tugas melaksanakan kegiatan litbang desain konseptual reaktor kogenerasi dengan tingkat daya menengah yang berpendingin gas helium dengan daya 200 MWt. Desain HTGR200K merupakan salah satu sistem pembangkit energi yang memiliki efisiensi energi paling besar, dan tingkat keselamatan inheren yang tinggi dan bersih. Komposisi geometri dan struktur teras didesain agar dapat menghasilkan keluaran pendingin gas helium bertemperatur 950 0C sehingga dapat digunakan untuk produksi hidrogen dan atau unit industri proses lainnya secara kogeneratif. Luaran gas helium bertemperatur sangat tinggi ini akan menimbulkan tegangan termal pada bola bahan bakar yang mengancam integritas sistem pengungkungan produk fisi di dalamnya. Oleh karena itu perlu dilakukan evaluasi karakteristika termal flow untuk menentukan distribusi temperatur bahan bakar bola dan outlet temperatur pendingin gas helium teras HTGR. Hal ini dilakukan dengan menggunakan modul Thermix-Konvek yang terintegrasi dalam program VSOP’94. Geometri teras HTGR dikerjakan dalam modul BIRGIT untuk model teras 2-D (R-Z dengan 5 kanal aliran pebble dalam teras aktif arah radial. Hasil evaluasi menunjukkan bahwa nilai tertinggi dan terendah temperatur yang terdapat pada teras   adalah sebesar 999.3 °C dan 886,5 °C. Demikian pula hasil temperatur tertinggi bahan bakar TRISO dan bahan bakar pebble di dalam teras, yaitu diperoleh sebesar  1510,20°C yang terletak pada lapisan bahan bakar inti UO2, di posisi z= 335.51 cm dan  r=0 cm. Analysis di lakukan pada laju massa aliran pendingin, tekanan dan daya masing-masing sebesar 120 kg/s, 7 Mpa dan 200MWth. Hasil

  20. Experimental operation of the RA reactor with 4 fuel channels containing 80% enriched dispersion fuel - Operational Report; Radni izvestaj - Eksperimentalna kampanja reaktora RA sa 4 kanala sa 80% obogacenim disperzionim gorivom

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Milosevic, M; Cupac, S; Kozomara, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-12-15

    Start of utilization of the new 80% enriched dispersion nuclear fuel is underway in the RA reactor core. Both economic and technical analyses were in favor of introducing the new fuel elements gradually into the RA reactor core. Thus overall theoretical and experimental analyses as well as other preparations are directed to transition regime based on gradual introducing of new fuel into the core, i.e. reactor core with two types of fuel. The objective of these analyses and preparation is establishment of conditions for safe reactor operation during transition period. The analyses and preparations are almost completed. The experimental data about fuel burnup during a time period of operation at nominal power i.e. daily decrease of excess reactivity is missing. This data is needed for planning the refueling (quantity of fresh fuel and frequency of refueling) during the transient period. This data can be obtained only by normal operation of the reactor during a period of time significantly longer than the period of attaining equilibrium poisoning, as time between two D{sub 2}O condensate overflows into the RA reactor core. Thus a ten day experimental campaign was planned to be done in December 1976. This report presents the most important results of safety analyses and preparation which show that, during this experimental period, the reactor operation is absolutely safe taking into account the most important parameters influencing reactor safety, as reactivity, thermal and temperature limits for fuel and the reactor, etc. Data to be obtained during this experimental campaign are significant because they would enable definition of future supply of fresh fuel during the transition period. [Serbo-Croat] Predstoji pocetak koriscenja novog 80% obogacenog uranskog disperzionog goriva u reaktoru RA. Ekonomske i tehnicke analize dale su prednost postupku postepenog uvodjenja novog goriva u reaktor RA. Prema tome, obimne teorijske i eksperimentalne analize i druge pripreme

  1. Professional Nuclear Materials Management; Gestion Industrielle des Matieres Nucleaires; Obrashchenie s yadernymi materialami na professional'nom urovne; Administracion Eficiente de Materiales Nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Forcella, A. A.; O' Leary, W. J. [Allis-Chalmers Manufacturing Company, Bethesda, MD (United States)

    1966-02-15

    manipulacion y su transporte resulten economicos y sea facil sustituirlos. La programacion de las operaciones de fabricacion debera reducir al mfnimo las perdidas de ingresos debidas a un almacenamiento improductivo excesivamente largo de materiales de costo elevado. Para la etapa de su empleo en el reactor, sera necesario un programa adecuado de redistribucion de los elementos combustibles a fin de lograr el grado maximo de combustion del material fisionable. Ademas, habra que reducir en la medida de lo posible los paros improductivos del reactor debidos a la redistribucion del combustible, a las inspecciones, etc. Cuando el combustible haya llegado a un grado predeterminado de agotamiento, el administrador tendra que disponer la forma mas economica de regeneracion de los materiales fisionables y de recuperacion de los subproductos. La administracion de materiales nucleares es por consiguiente un factor esencial para la reduccion de los costos del ciclo del combustible y para lograr que el costo unitario de la energia producida resulte economicamente interesante. (author) [Russian] Daets ja opisanie ob{sup e}ma rabot na tipichnom jenergetichesko m reaktore v SShA. Poskol'ku jetot reaktor finansiruetsja chastnym kapitalom, odna iz osnovnyh objazannostej operatora sostoit v obespechenii sohrannosti kapitalovlozhenij i poluchenii opredelennoj pribyli. Vvidu bol'shoj sebestoimosti jadernyh materialov neobhodimo postojanno obespechivat' nadlezhashhuju bezopasnost' i uchet s cel'ju svedenija k minimumu poter', ne svjazannyh s obespe- cheniem bezopasnosti i ucheta jadernyh materialov. Neobhodimo vse tshhatel'no produmyvat' i planirovat' zaranee, chtoby izbegat' nenuzhnyh zatrat i rashodovanija kapitala ili snizhe- nija razmerov pribyli v rjade oblastej. Pojetomu administrator, vedajushhij jadernymi materialami, dolzhen zaranee uchityvat' vse nepredvidennye obstojatel'stva i osushhestvljat' postojannyj kontrol' nad otklonenijami ot standartov pri planirovanii vo vremja a

  2. RA reactor operation in 1991, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1991. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1992-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1991, three major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. Renewal of the reactor instrumentation was started but but it is behind the schedule in 1991 because the delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA for political reasons. In 1991 most of the existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Training of the existing personnel was done regularly, but lack of financial support prevented employment of new workers that would be needed for operation in shifts and regular maintenance. [Serbo-Croat] U proteklom periodu reaktor RA nije bio u pogonu zato sto je 30. jula Republicki komitet za zdravlje i socijalnu politiku republike Srbije, zabranio njegov rad zbog toga sto reaktor ne poseduje sistem za udesno hladjenje i ne poseduje odgovarajuce filtere u sistemu specijalne ventilacije. Kako bi se ubuduce mogao obezbediti

  3. KETAHANAN KOROSI PADUAN Al-Mg 5052 DI DALAM AIR PENDINGIN NETRAL MENGANDUNG KLORIDA

    Directory of Open Access Journals (Sweden)

    Dicky Tri Jatmiko

    2015-07-01

    Full Text Available KETAHANAN KOROSI PADUAN Al-Mg 5052 DI DALAM AIR PENDINGIN NETRAL MENGANDUNG KLORIDA. Paduan Al-Mg 5052 adalah material yang biasa digunakan untuk kelongsong elemen bakar nuklir karena serapan fluks netronnya rendah dan tahan korosi di dalam air demineralisasi pada kondisi operasi reaktor. Makalah ini difokuskan untuk mengetahui ketahanan korosi paduan Al-Mg 5052 di dalam air dengan pH netral dan mengandung klorida sebagai pengganti air demineralisasi pendingin primer Reaktor Serba Guna GA Siwabessy (RSG-GAS. Penelitian mencakup pengukuran laju korosi menggunakan metode Tafel, prediksi mekanisme korosi menggunakan metode voltametri siklik dan analisa produk korosi dengan metode difraksi sinar X. Percobaan dilakukan dengan variasi temperatur 30°C, 35°C, 40°C, dan 45°C, serta variasi konsentrasi larutan natrium  klorida 0,05 M, 0,25 M, dan 0,5 M. Hasil penelitian ini menunjukkan bahwa paduan Al-Mg 5052 terkorosi dengan kategori “dapat diabaikan” hingga “sedang” dalam larutan natrium klorida menjadi produk yang larut dalam air pada satu tahap reaksi oksidasi irreversible.   CORROSION RESISTANCE OF Al-Mg ALLOY 5052 IN CHLORIDE CONTAINING NEUTRAL COOLING WATER. Al-Mg alloy 5052 is a material used as nuclear fuel element cladding due to its low neutron flux absorption and high corrosion resistance in demineralized water. This research is focused to know of the corrosion resistance of Al-Mg alloy 5052 in chloride containing neutral water used as demineralized primary cooling water substitute in GA Siwabessy Multi Purpose Reactor (RSG-GAS. This research covers the corrosion rate measurement using the Tafel method, corrosion process prediction using cyclic voltammetry method and corrosion product analysis using X-Ray Diffraction method. The experiments are carried out at temperature variation of 30°C, 35°C, 40°C and 45°C, as well as sodium chloride concentration of 0.05 M, 0.25 M and 0.5 M. The research results show that Al-Mg alloy 5052

  4. Influence of dissolved product gas on organism retention in biogas tower reactors; Der Einfluss geloester Produktgase auf den Organismenrueckhalt in Biogas-Turmreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Pietsch, T.; Maerkl, H. [Technische Univ. Hamburg-Harburg, Hamburg (Germany). Arbeitsbereich Bioprozess- und Bioverfahrenstechnik

    1999-07-01

    In biogas tower reactors, considerable oversaturations of CO{sub 2} dissolved in molecular form in the liquid phase can occur, compared to the thermodynamic steady state with the gas phase. In buildings of low height, upflow designs cause biological CO{sub 2} production along the reactor to saturate the liquid phase with carbonic acid, and also cause the pH value increasing from acid degradation to bind CO{sub 2} in the form of hydrogen carbonate HCO{sup -}{sub 3}. Where buildings are very high, the liquid phase becomes degassed through a decrease in CO{sub 2} partial pressure because of decreasing hydrostatic pressure along the length of the reactor. Rising gas bubbles in the liquid phase as well as enclosed gas bubbles in biomass particles slow down their sedimentation considerably and can result in flotation of biomass particles owing to gas expansion from declining hydrostatic pressure. A sedimentation characteristics for biomass under decreasing hydrostatic pressure is given. Conditions critical to biomass retention are energy input into CO{sub 2}-oversaturated liquids as well as dynamically rapid drops in pH value owing to associated CO{sub 2} degassing. (orig.) [German] In Biogas-Turmreaktoren koennen erhebliche Uebersaettigungen von molekular geloestem CO{sub 2} in der Fluessigphase gegenueber dem thermodynamischen Gleichgewichtszustand mit der Gasphase auftreten. Bei geringer Bauhoehe fuehrt bei upflow-Konzepten die biologische CO{sub 2}-Produktion entlang des Reaktors zu einer Aufsaettigung der Fluessigphase mit Kohlensaeure und der durch Saeureabbau ansteigende pH-Wert zu einer Bindung des CO{sub 2} in Form des Hydrogencarbonats HCO{sub 3}{sup -}. Sehr grosse Bauhoehen fuehren zu einer Entgasung der Fluessigphase durch Abnahme des CO{sub 2}-Partialdruckes aufgrund des abnehmenden hydrostatischen Druckes entlang der Reaktorhoehe. Aufsteigende Gasblasen in der Fluessigphase sowie eingeschlossene Gasblasen in Biomassepartikeln mindern deren

  5. Operation, maintenance and utilization of the RA reactor, Annual report for 1980; Pogon, odrzavanje i eksploatacija reaktora RA, Izvestaj o radu u 1980. godini

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1980-12-15

    During 1980 the activities of RA reactor staff was conducted in two directions: repair works and determination of the state of the existing equipment; and preparing and constructing additional equipment and completing the documentation needed to fulfill the new legal regulations. preparation of the existing equipment for future operation was finished, but construction of the additional equipment (according to the regulations) is not done at planned rate due to different reasons such as lack of budget, administration, import. This report includes a chapter devoted to the analysis of the operation difficulties that caused prohibition of reactor operation as well as difficulties in achievement of the planned activities for 1980. Data about financial issues are included as well. [Serbo-Croat] U toku 1980. godine rad OOUR-a Nuklearni reaktor RA odvijao se u dva smera: remontnim radovima i utvrdjivanju stanja postojece opreme i u pripremi i ugradnji potrebne dodatne opreme i kompletiranju neophodne dokumentacije prema novim zakonskim propisima. Priprema postojece opreme za rad je u potpunosti izvrsena, dok se projekti i ugradnja dodatne opreme (prema zakonskim obavezama) ne odvija potrebnom brzinom zbog raznih objektivnih razloga, kao nedostatak sredstava, odobrenja, uvoza. Ovaj izvestaj sadrzi poglavlje posveceno analizi teskoca u radu koje su dovele do zabrane rada reaktora, i teskoca u realizovanju plana rada u 1980. godini, ako i podatke o finansiranju projekta 'pogon, odrzavanje i eksploatacija reaktora RA'.

  6. Pengaruh Aerasi dan Sumber Nutrien terhadap Kemampuan Alga Filum Chlorophyta dalam Menyerap Karbon (Carbon Sink untuk Mengurangi Emisi CO2 di Kawasan Perkotaan

    Directory of Open Access Journals (Sweden)

    Lancur Setoaji

    2013-09-01

    Full Text Available Penelitian terkait mitigasi pemanasan global, khususnya dalam penyerapan karbon dioksida (CO2, menjadi fokus utama di kalangan ilmuwan dunia. Secara alamiah, karbon dioksida dapat diserap oleh tumbuhan hijau, laut, karbonasi batuan kapur, dan alga. Pigmen hijau dalam alga atau klorofil dapat menyerap karbon dioksida dalam proses fotosintesis. Alga memiliki pertumbuhan yang sangat cepat sehingga cocok digunakan sebagai carbon sink. Penelitian terkait carbon sink ini bertujuan untuk menentukan kemampuan rata-rata serapan CO2 oleh alga di kawasan perkotaan dan menentukan pengaruh aerasi dan variasi sumber N terhadap pertumbuhan dan perkembangan alga. Penelitian ini dilakukan dalam skala laboratorium menggunakan reaktor dengan proses batch. Sampel alga yang digunakan didapatkan dari hasil pengembangbiakan yang bersumber dari perairan di kawasan perkotaan. Penelitian ini menggunakan dua variabel uji, yaitu aerasi dan sumber nutrien. Jumlah karbon dioksida yang diserap didapatkan dari perbandingan stoikiometri pada reaksi fotosintesis.  Berdasarkan perbandingan stoikiometri tersebut diketahui bahwa 1 gram sel alga yang terbentuk sebanding dengan 1,92 gram CO2 yang diserap. Dari hasil penelitian, alga dengan penambahan pupuk urea dapat menyerap 4,87 mg CO2/hari dalam kondisi tanpa aerasi atau 3,84 mg CO2/hari dengan aerasi. Sedangkan alga dengan penambahan pupuk NPK dapat menyerap 3,61 mg CO2/hari dalam kondisi tanpa aerasi atau 3,01 mg CO2/hari dengan aerasi.

  7. KONVERSI KATALITIK MINYAK SAWIT UNTUK MENGHASILKAN BIOFUEL MENGGUNAKAN SILIKA ALUMINA DAN HZSM-5 SINTESIS

    Directory of Open Access Journals (Sweden)

    Nurjannah Nurjannah

    2012-02-01

    Full Text Available Terbatasnya sumber energi fosil menyebabkan perlunya pengembangan energi terbarukan yang berasal dari alam dan dapat diperbaharui. Penggunaan bahan bakar minyak bumi, baik dari penggunaan berupa alat transportasi maupun dari penggunaan oleh industri sangat mencemari lingkungan karena tingkat polusi yang ditimbulkan sangat tinggi sehingga perlu mencari bahan bakar alternatif pengganti bahan bakar gasoline, solar, dan kerosene dari minyak nabati. Penelitian dilakukan dalam dua tahapan yaitu sintesa katalis dan proses katalitik cracking. Silika alumina disintesa menggunakan metode Latourette dan HZSM-5 disintesa menggunakan metode Plank. Hasil sintesa dikarakterisasi dengan Penyerapan Spektroskopi Atomis (AAS menunjukkan bahwa silika alumina dan HZSM-5 mempunyai Si/Al 198 dan 243. Luas permukaan  silika alumina dan HZSM-5 diperoleh dari analisa Brunauer Emmet Teller (BET yaitu 149,91-213,35 m2.g-1 dan ukuran pori rata-rata adalah 13oA. Perengkahan katalitik dilakukan dalam suatu mikroreaktor fixed bed pada temperatur 350-500°C dan laju alir gas N2 100-160 ml.min-1 selama 120 min. Hasil perengkahan dianalisa dengan metode gas kromatografi. Hasil yang diperoleh untuk katalis HZSM-5 fraksi gasoline dengan yield tertinggi 28,87%, kerosene 16,70%, dan diesel 12,20%  pada suhu reaktor 4500C dan laju gas N2 100 ml/menit.

  8. Operational experience gained with the failed fuel rod detection system in nuclear power plants

    International Nuclear Information System (INIS)

    Boehm, H.H.; Forch, H.

    1985-01-01

    Brown Boveri Reaktor GmbH together with Krautkramer Company developed such a FAILED FUEL ROD DETECTION SYSTEM (FFRDS) which allows to located defective fuel rods without dismantling the fuel assembly or pulling of individual rods. Since 1979 the FFRDS is employed successfully in various nuclear power plants in Europe, USA, Japan, and Korea. The short inspection time and the high reliability of the method make the FFRDS a true competitor to the sipping method. In this paper the authors discuss the method and the design of the system, the equipment set-up, its features and the experience gained so far. The system has been performed and automated to such an extent that within a short installation period series of fuel assemblies can be tested with relatively short intervals of time (5 minutes for BWR and 7 minutes for PWR fuel assemblies per side). The ability of the system for deployment under various conditions and the experience gained during the past six years have made this system universally applicable and highly sensitive to the requirements of NDT during outages and for transport of FAs to intermediate storage facilities. Comparison of FFRDS to conventional sipping has indicated in several instances that the FFRDS is superior to the latter technique

  9. Thermal-Hydraulic Experiment To Test The Stable Operation Of A PIUS Type Reactor

    International Nuclear Information System (INIS)

    Irianto, Djoko; Kanji, T.; Kukita, Y.

    1996-01-01

    An advanced type of reaktor concept as the Process Inherent Ultimate Safety (PIUS) reactor was based on intrinsically passive safety considerations. The stable operation of a PIUS type reactor is based on the automation of circulation pump speed. An automatic circulation pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed the PIUS-type reactor. In principle this control system maintains the fluid temperature at the axial center of the lower density lock at average of the fluid temperatures below and above the lower density lock. This control system will prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments using atmospheric pressure thermal-hydraulic test loop which simulated the PIUS principle. The experiments such as: start-up and power ramping tests for normal operation simulation and loss of feedwater test for an accident condition simulation, carried out in JAERI

  10. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    1986-11-01

    RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

  11. Hydrogen Mixing Studies (HMS) assessment manual

    International Nuclear Information System (INIS)

    Lam, K.L.; Wilson, T.L.; Travis, J.R.

    1993-06-01

    This report documents some calculations performed to assess the Hydrogen Mixing Studies (HMS) code. Results are presented first for some analytical test problems, including laminar flow and mass diffusion. The von Karman vortex street problem and the Sandia FLAME Facility and Heiss Dampf Reaktor (HDR) containment facility test problems are then discussed. For the analytical problems, the code gave results that agree exceptionally well with the analytical solutions. Calculations for the von Karman vortex street problem were performed at selected Reynolds numbers for several obstacle types. The computed flow patterns agree well with experimental observations-specifically the occurrence of a vortex street (double row of vortices) above a critical Reynolds number. Calculations for the von Karman vortex street problem were performed at selected Reynolds numbers for several obstacle types. The computed flow patterns agree well with experimental observations-specifically the occurrence of a vortex street (double row of vortices) above a critical Reynolds number. The last assessment problem involves modeling the experiment T31.5. The experiment was carried out in the HDR containment building, which is a large, multi-compartment facility (11 300 m 3 free volume in 72 compartments). In the experiment, a steam-water mixture was first injected into the containment to simulate a large-break blowdown of a pressure vessel, and then superheated steam was injected that was followed by a release of helium-hydrogen light gas. The calculated results (pressure, temperature, and gas concentrations) agree reasonably well with the experimental data

  12. The Quality Standard 50-SG-Q8 Application On Research And Development Activities In P2TRR

    International Nuclear Information System (INIS)

    Nababan, N.; Rofei, A.

    2000-01-01

    Utilization of nuclear reactor installation should reference safety, Environment, and instalation requrement, and shoul incorporate user requirement. The requirement are adequately addressed in the IAEA standar No. 50-sg-Q8 for R and D work in nuclear instalation. A study to this standar has been carried out and it is found that there are three important factor of QA methodologi necessary to be aplied for R and D activities. They are management responsibilities, performance of the R and D works, and carrying out assesment over the R and D project. The R and D activities performing in Centre for Recearch Reaktor Technology Development (P2TRR) shoul meet the requirement in the R and D program assigned by Batan management, Such as Strastegy and Policy (jakarta), Strategic Plan (Restra), and Five-Year Plan. A study to this program also has been carried out and it is found that QA methodologies are necessary to be aplied, Especially during the data gathering activities, and its contribution to the safety function or as input to actual nuclear reactor installation or systems

  13. PEMBUATAN KALSIUM KARBONAT DARI BITTERN DAN GAS KARBON DIOKSIDA SECARA KONTINYU

    Directory of Open Access Journals (Sweden)

    Soemargono Soemargono

    2012-01-01

    Full Text Available Kalsium karbonat  yang  digunakan  dalam  industri- industri cat, karet, dan  kertas  harus  mempunyai  mutu yang  tinggi, terutama  kemurnian  dan kehalusannya.Untuk itu, Indonesia masih  mendatangkan  kalsium  karbonat murni dari luar negeri dalam jumlah yang cukup besar. Bittern merupakan bahan buangan industri garam yang disebut juga air tua, mengandung senyawa kalsium. Karbon dioksida biasanya berasal dari hasil pembakaran yang masuk ke udara. Kandungannya di udara kecil, tetapi berpotensi sebagai pencemar. Dengan mereaksikan kalsium yang terkandung dalam bittern dengan gas CO2 akan terbentuk CaCO3 dalam suasana basa. Pembentukan kalsium karbonat dilakukan dengan proses kontinyu dalam reaktor kolom bersekat miring. Dari hasil penelitian diperoleh bahwa pengendapan magnesium dengan larutan ammonia menyebabkan kandungan kalsium ikut terdegradasi. Hasil terbaik yang diperoleh dicapai pada kondisi pH awal, kecepatan alir gas CO2, kecepatan alir cairan, dan suhu masing-masing pada 8,7; 2265 mL/menit; 10 mL/menit; dan 303 K, dengan konversi sebesar 38,40%. Produk berupa CaCO3, yang diperoleh mempunyai kemurnian sebesar 21,34%.

  14. Structural analysis of aircraft impact on a nuclear powered ship

    International Nuclear Information System (INIS)

    Dietrich, R.

    1976-01-01

    As part of a general safety analysis, the reliability against structural damage due to an aircraft crash on a nuclear powered ship is evaluated. This structural analysis is an aid in safety design. It is assumed that a Phantom military jet-fighter hits a nuclear powered ship. The total reaction force due to such an aircraft impact on a rigid barrier is specified in the guidelines of the Reaktor-Sicherheitskommission (German Safety Advisory Committee) for pressurized water reactors. This paper investigates the aircraft impact on the collision barrier at the side of the ship. The aircraft impact on top of the reactor hatchway is investigated by another analysis. It appears that the most unfavorable angle of impact is always normal to the surface of the collision barrier. Consequently, only normal impact will be considered here. For the specific case of an aircraft striking a nuclear powered ship, the following two effects are considered: Local penetration and dynamic response of the structure. (Auth.)

  15. Investigation of corrosion of materials of the irradiation device in the RA reactor; Ispitivanje korozije materijala uredjaja za ozracivanje na reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Zaric, M; Mance, A; Vlajic, M [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    Devices for sample irradiation in the vertical RA reactor channels will be made of aluminium alloys. According to the regulations concerned with introducing materials into the RA reactor core, corrosion characterisation of these materials is an obligation. Corrosion properties of four aluminium alloys were investigated both in contact with stainless steel and without it. First part of this report deals with the corrosion testing of aluminium alloys in water by gravimetric and electrochemical methods. Bi-distilled water at temperatures less than 100 deg C was used. Second part is related to aluminium alloys corrosion in carbon dioxide gas under experimental conditions. The second part of research was initiated by the design of the head of the independent CO{sub 2} loop for samples cooling. [Serbo-Croat] Uredjaji za ozracivanje u vertikalnim kanalima reaktora RA, bice napravljeni od legura aluminjuma. Prema propisima o unosenju materijala u RA reaktor materijali se moraju prethodno ispitati i sa stanovista korozije. Ispitivane su korozione pojave na cetiri aluminjumske legure sa i bez kontakta sa nerdjajucim celikom. Prvi deo ovog rada tretira pitanje korozije legura aluminijuma u vodi gravimetrijskim i elektrohemijskim metodama. Koriscena je bidestilovana voda na temperaturi do 100 deg C. Drugi deo se odnosi na ispitivanje ponasanja legura aluminijuma u gasovitom ugljen dioksidu pod uslovima eksperimenta. Drugi deo istrazivanja izvrsen je za potrebe izgradnje glave petlje nezavisnog kola za hladjenje uzoraka gasovitim CO{sub 2}.

  16. PENGELOLAAN LIMBAH PETERNAKAN SAPI UNTUK MENINGKATKAN KAPASITAS PRODUKSI PADA KELOMPOK TERNAK PATRA SUTERA

    Directory of Open Access Journals (Sweden)

    Danang Dwi Saputro

    2014-02-01

    Full Text Available Kelompok ternak Patra Sutera di Desa Ledok Kecamatan Sambong Kabupaten Blora yang berdiri pada tahun 2013 telah mempunyai sapi 8 ekor yang berada di kandang komunal yang dikelola oleh 6 anggota kelompok. Dalam satu hari setiap ekor sapi dapat menghasilkan limbah padat sebanyak 20-30 kg dan limbah cair sebanyak 100-150 liter yang selama ini belum dikelola dengan baik. Limbah dari kegiatan ternak belum terolah dengan baik dan dibuang ke lingkungan sehingga menimbulkan dampak negatif bagi kesehatan masyarakat sekitar kandang. Salah satu cara untuk mengatasi kondisi ini adalah dengan memberikan pelatihan keterampilan atau pendampingan bagaimanakah teknik pembuatan pupuk organic dan reaktor biogas sederhana, mengoperasikan, serta memanfaatkan gas yang dihasilkan. Dalam kegiatan ini akan diberikan pelatihan keterampilan bagaimana cara mengolah limbah ternah untuk dijadikan pupuk dan pestisida organik,serta pengelolaan biodigester. Dari kegiatan ini Anggota kelompok ternak Patra Sutera mendapat pengetahuan dan mengolah limbah kotoran ternak (padat dan cair yang keliar dari biodigester menjadi pupuk yang lebih bermanfaat.

  17. PERANCANGAN PEMBANGKIT LISTRIK TENAGA SAMPAH ORGANIK ZERO WASTE DI KABUPATEN TEGAL (STUDI KASUS DI TPA PENUJAH KABUPATEN TEGAL

    Directory of Open Access Journals (Sweden)

    Abdul Muiz Liddinillah Sanfiyan

    2017-12-01

    Full Text Available Permasalahan sistem pengolahan sampah yang ada di Kabupaten Tegal adalah masih menggunakan sistem Open Dumping. Berdasarkan data yang diperoleh dari Badan Pusat Statistik (BPS Kabupaten Tegal pada tahun 2016, komposisi sampah organik adalah yang terbesar kedua setelah sampah plastik dan sangat berpotensi mengalami penambahan setiap tahunnya. Tujuan dari penelitian ini adalah untuk membuat perancangan pembangkit listrik tenaga sampah organik zero waste di Kabupaten Tegal, dengan studi kasus di Tempat Pembuangan Akhir (TPA sampah Penujah. Objek dalam penelitian ini adalah sistem pengolahan sampah organik yang ada di Kabupaten Tegal dengan menggunakan sistem pengolahan sampah zero waste. Sistem pengolahan sampah organik zero waste adalah sistem pengolahan sampah yang tidak menghasilkan sampah kembali. Jadi, diharapkan jumlah sampah organik akan berkurang secara bertahap. Sampah organik dapat dirubah menjadi biogas melalui proses fermentasi yang dibantu oleh bakteri secara anaerob di dalam reaktor biodigester. Biogas tersebut ditampung di dalam tempat penampungan untuk kemudian didistribusikan ke dalam genset biogas sebagai bahan bakar pembangkit listrik. Sisa pengolahan biogas dapat dirubah menjadi pupuk cair dan pupuk kompos yang bernilai ekonomis.

  18. Research reactor RB, technical characteristics and experimental possibilities; Zbornik radova, Konferencija o koriscenju nuklearnih reaktora u Jugoslaviji

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Vranic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1978-05-15

    Nuclear research reactor RB tn the Nuclear Engineering Laboratory at the Institute of Nuclear Sciences 'Boris Kidric' in Vinca is the first reactor system built in Yugoslavia in 1958. In this report, the basic technical characteristics of this reactor are described, as well as the experimental possibilities it offers to the users. Its relatively simple construction and flexibility enables direct measurements of a series of physical parameters, and the absence of the biological protection shield makes it very useful for Various biological and other irradiations and dosimetric measurements Where strong neutron source is required. (author) Istrazivacki nuklearni reaktor RB u Laboratoriji za nuklearnu energetiku i tehnicku fiziku Instituta za nuklearne nauke 'Boris Kidric' u Vinci je prvi reaktorski sistem izgradjen u Jugooslaviji 1958. godine. U ovom radu opisane su osnovne tehnicke karakteristike tog reaktora, kao i mogucnosti za izvodjenje eksperimenata koje on pruza korisnicima. Njegova relativno jednostavna konstrukcija i fleksibilnost omogucavaju da se na njemu izvrse direktna merenja niza fizickih parametara, a s druge strane odsustvo bioloskog zastitnog omotaca cini ga veoma pogodnim za razna bioloska i druga ozracivanja, a takodje i dozimetrijska merenja gde se zahteva snazan izvor neutrona. (author)

  19. Parní generátor reaktoru ESFR

    OpenAIRE

    Bátěk, David

    2012-01-01

    Tato diplomová práce se zabývá parním generátorem pro reaktory ESFR (Evropský Rychlý Sodíkem Chlazený Reaktor), jenž je vyhříván tekutým sodíkem. V úvodních kapitolách jsou teoretické informace o parametrech ESFR a porovnání s výměníky tepla v jaderných elektrárnách pracujících na stejném principu (sodík jako chladivo). Následuje návrhová část, kam patří úvod do problematiky výpočtu, volba materiálu a koncepce výměníku a samotná výpočtová část, která zahrnuje tepelný, hydraulický a pevnostní ...

  20. Nuclear fuel in water reactors: Manufacturing technology, operational experience and development objectives in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Holzer, R.; Knoedler, D.

    1977-01-01

    The nuclear fuel industry in the Federal Republic of Germany comprises the full range of manufacturing capabilities for pressurized-, boiling- and heavy-water reactor technology. The existing manufacturing companies are Reaktor-Brennelement Union (RBU) and Alkem. RBU makes natural and enriched UO 2 -fuel assemblies, starting with powder preparation. Facilities to produce UO 2 -gadolinia and UO 2 -ThO 2 fuel are also available. Alkem manufactures mixed-oxide UO 2 /PuO 2 fuel and fuel rods. Zircaloy cladding tubes are produced by Nuklearrohr-Gesellschaft (NRG) and Mannesmannroehren-Werke (MRW). Construction of a new fuel manufacturing plant has been announced by Exxon. Supplementary to quality control, an integrated quality assurance system has been established between the reactor vendor's fuel design and engineering division and the existing manufacturing companies for fuel and tubing. Operating experience with LWR and HWR fuel dates back to 1964/65 and has shown good performance. Possible reasons for a small fraction of defective rods could be identified quickly by a fast feedback system incorporating close co-operation between Kraftwerk Union (KWU) and the utilities. KWU combines fuel development, hot-cell and pool-side service facilities as well as fuel technology linked to manufacturing. The responsibility of KWU for core and fuel design, which enabled an integral optimization, was also an important reason for the successful operation and design flexibility. (author)

  1. Data base formation for important components of reactor TRIGA MARK II

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, R; Mavko, B; Kozuh, M [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [Slovenian] V referatu smo prikazali raziskavo, v okviru katere smo za raziskovalni reaktor TRIGA MARK II v Podgorici izoblikovali specificno bazo podatkov. Zbrali smo podatke obratovanja reaktorja od leta 1985 do 1990. Rezultate raziskave dogodkov smo razdelili v dve glavni skupini. V prvo spadajo obratovalni podatki o komponentah, v drugo skupino pa spadajo zagoni oz. zaustavitve reaktorja. Podatke smo ovrednotili z modelom v casovnem prostoru in z modelom na zahtevo. Parametre modelov smo dolocili s klasicno metodo. Opisane baze podatkov so uporabne povsod, kjer so lahko posledice nezanesljivega delovanja sistemov velike. [author].

  2. Remerschen nuclear power station with BBR pressurized water reactor

    International Nuclear Information System (INIS)

    Hoffmann, J.P.

    1975-01-01

    On the basis of many decades of successful cooperation in the electricity supply sector with the German RWE utility, the Grand Duchy of Luxemburg and RWE jointly founded Societe Luxembourgeoise d'Energie Nucleaire S.A. (SENU) in 1974 in which each of the partners holds a fifty percent interest. SENU is responsible for planning, building and operating this nuclear power station. Following an international invitation for bids on the delivery and turnkey construction of a nuclear power station, the consortium of the German companies of Brown, Boveri and Cie. AG (BBC), Babcock - Brown Boveri Reaktor GmbH (BBR) and Hochtief AG (HT) received a letter of intent for the purchase of a 1,300 MW nuclear power station equipped with a pressurized water reactor. The 1,300 MW station of Remerschen will be largely identical with the Muelheim-Kaerlich plant under construction by the same consortium near Coblence on the River Rhine since early 1975. According to present scheduling, the Remerschen nuclear power station could start operation in 1981. (orig.) [de

  3. Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments

    Science.gov (United States)

    Tregoures, Nicolas; Bandini, Giacomino; Foucher, Laurent; Fleurot, Joëlle; Meloni, Paride

    The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MWe PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.

  4. Slovak brown coals as a feedstock for the active coke production

    Directory of Open Access Journals (Sweden)

    Sobolewski Aleksander

    1998-09-01

    Full Text Available V èlánku sa venuje pozornos možnostiam výroby aktívneho koksu zo slovenského hnedého uhlia Bane Cíge¾. Príprava aktívneho koksu bola uskutoènená v laboratórnych podmienkach a v poloprevádzke. Surové uhlie sa podrobilo vysokoteplotnej pyrolýze v retorte s pevným roštom, v klasickej rotaènej peci, ako aj vo fluidnom reaktore. Následne boli urèované adsorpèné charakteristiky získaného aktívneho koksu, ktoré sa porovnávali s charakteristikami komerène vyrábaného koksu. V príspevku sa taktiež diskutujú možnosti aplikácie pripraveného aktívneho koksu v technológiách ochrany životného prostredia.

  5. Neutron Diffractometer; Neutronski difraktometar

    Energy Technology Data Exchange (ETDEWEB)

    Zivadinovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    RA nuclear reactor is considered as a relatively strong neutron source producing the thermal neutron flux of about 3x10{sup 13} n/cm{sup 2} sec when operating at nominal power of 6.5 MW. Neutron diffraction method is applied in the field of solid state physics, material science, crystallography, magnetism, nuclear physic. Neutron diffractometer at the RA reactor consists of: system for obtaining collimated neutron beam from the horizontal experimental channel neutron monochromator; goniometer and electronic equipment for measurements and collecting the the measurement data. Nuklearni reaktor RA koji pri radu na snazi od 6,5 MW ima fluks termalnih neutrona oko 3x10{sup 13} n/cm{sup 2} sec predstavlja relativno jak izvor neutrona. Tehnika difrakcije neutrona primenjuje se u istrazivanjima fizike crvstog stanja, strukture materijala, kristalografije, magnetizma, nuklearne fizike. Neutronski difraktometar na reaktoru RA sastoji se od sistema za dobijanje kolimisanog snopa neutrona kroz horizontalni kanal reaktora; neutroskog monohromatora; goniometra i elektronskih uredjaja za merenja i registrovanje rezultata. Ovaj izvestaj sadrzi detaljan opis i seme neutronskog difraktometra sa pratecom opremom i elektronskim komponentama.

  6. Failed fuel rod detection system and computerized manipulator during outages

    International Nuclear Information System (INIS)

    Boehm, H.H.; Foerch, H.

    1984-01-01

    During regular outages spent fuel assemblies need to be replaced and relocated within the core. Defective fuel rods in particular fuel assemblies have to be removed from further service and before delivery of such faulty fuel assemblies to a reprocessing plant. The system which Brown Boveri Reaktor GmbH and Krautkraemer have developed in the Federal Republic of Germany is capable of directly locating the defective rods in a proper fuel assembly. Inspection times are comparable to those of standard sipping methods, with the advantages of immediately available results and direct identification of the defective fuel rods. During the repair of fuel assemblies this system allows withdrawal of individual defective rods. With the sipping method all the fuel rods of a defective fuel assembly need to be removed and inspected by eddy current testing. During steam generator inspection and repair personnel are exposed to ample radiation. A remotely controlled, computerized manipulator was used to significantly reduce the radiation dose by automating steps in the procedures; at the same time inspection and repair times were reduced. The main features of the manipulator are a rigid component construction of the leg and two arms, and a resolver control for horizontal and vertical motion that enables rapid and accurate access to a desired tube (author)

  7. PENGURANGAN KADAR CO2 MENGGUNAKAN SPIRULINA PLATENSIS DALAM TUBULAR BIOREACTOR

    Directory of Open Access Journals (Sweden)

    Zainal Syam Arifin

    2015-06-01

    Meningkatnya jumlah penduduk berdampak pada peningkatan kebutuhan energi. Di lain pihak, industri pembangkit energi dituding sebagai salah satu penyumbang karbon dioksida sekitar 25% dari total emisi CO2 di seluruh dunia. Disisi lain, produksi biogas yang bertujuan untuk mengatasi peningkatan kebutuhan energi justru menghasilkan karbon dioksida pada kisaran 25 – 50% volume. Untuk mengatasi hal ini, diperlukan metode yang murah, optimum dan efisien serta ramah lingkungan dalam mengurangi kadar CO2 dengan menggunakan spirulina platensis.  Penelitian ini bertujuan membuat model matematik dan menemukan kecepatan aliran yang optimum untuk menurunkan kadar CO2 dengan menggunakan Spirulina Platensis. Penelitian ini menggunakan reaktor tubularterbuat dari kaca (D = 2,6 cm pada suhu 30°C dan disinari dengan lampu TL Philips fluoresen 36 Watt, temperatur warna: 6.200K cool daylight, light output: 2.600 lm, 72 lm/W. Reaktor tubular ditempatkan dalam kotak yang dilapisi dinding dengan kertas perak pada ketiga sisinya. Dengan model matematik reaktor tubular, dapat diprediksi konstanta kecepatan reaksinya. Berdasarkan grafik hasil perhitungan data, kecepatan volumetrik optimumnya juga dapat diprediksi. Variasi flowrate yaitu 0,25 mL/detik, 0,35 mL/detik, 0,5 mL/detik, 0,75 mL/detik, 1 mL/detik. Sumber karbon adalah CO2 99,99%. Pengamatan pertumbuhan Spirulina dilakukan pada flow rate 0,25 mL/detik dengan kadar berat kering mula – mula 2,1208 g/L. Hasil penelitian ini menunjukkan bahwa aliran lambat (flowrate rendah merupakan cara yang lebih efektif dalam mengurangi karbon dioksida menggunakan spirulina platensis (= 2,82×10-4 detik-1. Nilai konversi tertinggi diperoleh pada kecepatan aliran volumetrik 0,25 mL/detik dan kecepatan optimumnya pada kisaran 0,3 – 0,4 mL/detik. Laju flux CO2 masuk sebaiknya kurang dari 0,047 mL/cm2.detik. Nilai Specific Growth Rate (µ Spirulina Platensis dalam penelitian ini yaitu 2,56×10-2 menit-1.   Kata Kunci: Spirulina platensis, reaktor tubular

  8. SAFIRA project B.3.3: in-situ-treatment of contaminated ground water by catalytic oxidation. Final report; Sanierungsforschung in regional kontaminierten Aquiferen (SAFIRA). Projekt B.3.3: In situ-Behandlung von kontaminierten Grundwaessern durch katalytische Oxidation. Teilvorhaben 1: Untersuchungen im Labormassstab. Teilvorhaben 2: Tests in der bench-scale-Anlage und Teilvorhaben 3: Die Erprobung in der Pilotanlage am Modellstandort. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, J.; Haentzschel, D.; Freier, U.; Wecks, M.

    2003-06-27

    A new technology for treatment of contaminated ground water was developed. In this process heterogeneous catalysts (full metal catalyst, mixed oxide catalyst or iron-containing zeolites) in combination with hydrogen peroxide are used. In the reactor catalytic oxidation and aerob biological degradation occur simultaneously. A complete degradation of chlorobenzene was observed in a bench-scale-equipment (2 liter) and also in the pilot plant at the model site located in Bitterfeld (30 liter reactor). The technology can be applied to the ground and waste water treatment. (orig.) [German] Fuer die Behandlung von Grundwaessern, die mit organischen Schadstoffen belastet sind, wurde ein neuartiges Verfahren entwickelt. Bei der katalytischen Oxidation werden heterogene Katalysatoren in Form von Vollmetall-, Mischoxid- und Traegerkatalysatoren in Verbindung mit Wasserstoffperoxid als Oxidationsmittel eingesetzt. In den Katalysereaktoren laufen die heterogen-katalytische Oxidation und der aerob-biologische Abbau nebeneinander ab. Es werden synergistische Effekte erzielt. Mit dem Verfahren wurde in einer bench-scale-Angle (2 Liter) und in der Pilotanlage am Modellstandort in Bitterfeld (30 l Reaktor) der Schadstoff Chlorbenzol vollstaendig umgesetzt. Das Verfahren kann zur Grund- und Abwasserbehandlung eingesetzt werden. (orig.)

  9. PENGOLAHAN LIMBAH CAIR INDUSTRI FARMASI FORMULASI DENGAN METODE ANAEROB-AEROB DAN ANAEROB-KOAGULASI

    Directory of Open Access Journals (Sweden)

    Farida Crisnaningtyas

    2016-05-01

    Full Text Available Studi ini membahas mengenai pengolahan limbah cair industri farmasi dalam skala laboratorium dengan menggunakan konsep anaerob-kimia-fisika dan anaerob-aerob. Proses anaerob dilakukan dengan menggunakan reaktor Upflow Anaerobic Sludge Bed reactor (UASBr pada kisaran OLR (Organic Loading Rate 0,5 – 2 kg COD/m3hari, yang didahului dengan proses aklimatisasi menggunakan substrat gula. Proses anaerob mampu memberikan efisiensi penurunan COD hingga 74%. Keluaran dari proses anaerob diolah lebih lanjut dengan menggunakan dua opsi proses: (1 fisika-kimia, dan (2 aerob. Koagulan alumunium sulfat dan flokulan kationik memberikan efisiensi penurunan COD tertinggi (73% pada kecepatan putaran masing-masing 100 rpm dan 40 rpm. Uji coba aerob dilakukan pada kisaran MLSS antara 4000-5000 mg/L dan mampu memberikan efisiensi penurunan COD hingga 97%. Hasil uji coba menunjukkan bahwa efisiensi penurunan COD total yang dapat dicapai dengan menggunakan teknologi anaerob-aerob adalah 97%, sedangkan kombinasi anaerob-koagulasi-flokulasi hanya mampu menurunkan COD total sebesar 72,53%. Berdasarkan hasil tersebut, kombinasi proses anaerob-aerob merupakan teknologi yang potensial untuk diaplikasikan dalam sistem pengolahan limbah cair industri farmasi. 

  10. Combination of biological treatment of waste air and liquid effluents in small and medium-sized businesses by using spherical carrier media in exchange with each other. Final report; Kombination der biologischen Behandlung von Abluft- und Abwasserreinigung in kleinen und mittleren Unternehmen durch die Verwendung von kugeligen Traegermaterialien im gegenseitigen Austausch. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Kunz, P.M.; Brunk, M.; Bentz, P.; Bach, K.; Marcos del Rio, O.; Stahl, B.; Wolf, C.

    2002-05-08

    A combined system for biological treatment of solvent-contaminated liquid and gaseous effluents was to be constructed. The system was to be mobile for demonstration purposes in industrial processes and was to serve for dimensioning of full-scale systems. The system works by a combined process via exchange of spherical carriers with a growth of micro-organisms. Two pilot systems of the Institut fuer Biologische Verfahrenstechnik, one for liquid and one for gaseous effluents, were combined for this purpose, ensuring exchange of carrier materials between the two units. (orig.) [German] Aufgabe im Rahmen dieses F+E-Vorhabens war es gewesen, eine kombinierte biologische Abluft- und Abwasserreinigungsanlage - speziell fuer loesungsmittelhaltige Abluft und Abwaesser - zu untersuchen. Dazu sollte eine Anlage aufgebaut werden, die auch als Demonstrations- und Versuchsanlage zu Industriebetrieben gebracht und dort zur Dimensionierung von full-scale-Anlage eingesetzt werden kann. Idee fuer dieses Vorhaben war gewesen, die Kombination beider Behandlungssysteme ueber den Austausch von mit Mikroorganismen bewachsenem Traegermaterial herzustellen. Am Institut fuer Biologische Verfahrenstechnik existierte bereits eine Pilotanlage fuer die biologische Abwasserreinigung mit kugeligem Traegermaterial (Vario-Bio-Reaktor) und eine Pilotanlage fuer die biologische Behandlung von Abluft nach dem Trickling-Filter-Prinzip mit ebenfalls kugeligem Traegermaterial. Diese beiden Anlagen sollten miteinander kombiniert und der Austausch des Traegermaterials hergestellt werden. (orig.)

  11. Operation and maintenance of the RA nuclear reactor for 1977, Report Annex I; Rad i iskoriscenost reaktora RA u 1977. godini, izvestaj Prilog I

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Stanic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1977-12-15

    RA reactor operation plan was fulfilled, meaning 28583 MWh. In addition to 183 days of operation at full power, during 1977 the reactor was operated for 14 days at lower power level according to the demand of users. The utilization level of rector irradiation capability (neutron flux and time of operation) was 14%. This annex includes detailed statistical data about reactor operation, utilization, power level, savings concerned with new 80% enriched fuel. All the 9 vertical experimental channels were used for irradiation in the reactor core. Digression from the action plan were caused by refueling and demand od the users. There have been 11 safety shutdowns, of which 6 caused by power cuts, 4 due to failure of the instruments, and 1 due to earthquake in March 1977. [Serbo-Croat] Planirani rad reaktora na nominalnoj snazi ostvaren je u iznosu od 28583 MWh. U toku 1977. godine reaktor je radio 14 dana na manjim snagama i u posebnom eksperimentalnom rezimu na zahtev korisnika. Iskoriscenost kapaciteta reaktora za ozracivanje uzoraka (na bazi neutronskog fluksa i vremena rada reaktora) bila je 41%. Ovaj izvestaj sadrzi detaljne statisticke podatke o radu i iskoriscenosti reaktora, podatke o ustedi goriva prelaskom na 80% obogaceno gorivo. Korisceno je svih 9 vertikalnih eksperimentalnih kanala u aktivnoj zoni. Uzroci odstupanja od plana rada osim zamene goriva bili su zahtevi korisnika. Bilo je 11 sigurnosnih zaustavljanja, 6 usled nestanka napona, 4 usled kvarova opreme i instrumentacije, i 1 put usled zemljotresa u martu.

  12. Investigations of coal ignition in a short-range flame burner using optical measuring systems; Untersuchungen zur Kohlezuendung am Flachflammenbrenner unter Verwendung optischer Messtechnik

    Energy Technology Data Exchange (ETDEWEB)

    Hackert, G.; Kremer, H.; Wirtz, S. [Bochum Univ. (Germany). Lehrstuhl fuer Energieanlagentechnik

    1999-09-01

    The short-range flame burner and the KOALA reactor of DMT are experimental facilities for realistic simulation of coal conversion processes at high temperatures and pressures in atmospheric conditions. The TOSCA system enable measurements of temperatures, sizes, shapes and velocities of the fuel particles, which serve as a basis for a three-dimensional simulation model of coal combustion. In the future, further parameter studies will deepen the present knowledge of coal dust combustion under pressure and enable optimisation of the numerical models for simulation of industrial-scale systems for coal dust combustion under pressure. [Deutsch] Mit dem Flachflammenbrenner und dem KOALA-Reaktor der DMT stehen Versuchsapparaturen zur Verfuegung, mit deren Hilfe die Kohleumwandlungsprozesse bei hohen Temperaturen unter Druck und unter atmosphaerischen Bedingungen realistisch wiedergegeben werden. Das TOSCA-System erlaubt dabei die Bestimmung von Temperaturen, Groessen, Formen und Geschwindigkeiten der Brennstoffpartikel. Diese Daten liefern die Grundlage fuer die Erstellung eines dreidimensionalen Simulationsmodells zur Modellierung der Kohleverbrennung. In Zukunft werden weitere Parameterstudien das Verstaendnis der Kohlenstaubdruckverbrennung vertiefen und ein Optimierung der numerischen Modelle ermoeglichen, so dass die Simulation grosstechnischer Kohlenstaubdruckverbrennungsanlagen realisiert werden kann. (orig.)

  13. Sintesis Furfural dari Bagas Tebu Via Reaksi Hidrolisa dengan Menggunakan Katalis Asam Asetat pada Kondisi Atmosferik

    Directory of Open Access Journals (Sweden)

    Nine Tria Rossa

    2015-12-01

    Full Text Available Telah dilakukan penelitian tentang hidrolisa bagas tebu menggunakan asam asetat sebagai katalis. Sebanyak 30 gram dihidrolisa dalam 300ml air yang mengandung katalis asam menggunakan  asetat sebesar 7 sampai 9% v/v dengan variabel waktu dan temperatur hidrolisa selama 1 sampai 4 jam dan 80oC sampai 103oC pada kondisi atmosferik menggunakan reaktor tipe batch. Penelitian ini bertujuan untuk mengetahui pengaruh temperatur dan waktu hidrolisa, serta konsentrasi katalis asam asetat terhadap perolehan furfural. Kemudian untuk menemukan kondisi paling efisien untuk memproduksi furfural menggunakan Response Surface Methodology (RSM dengan Software Design Expert 7.0.0. Hasil penelitian menunjukkan bahwa dengan penambahan waktu dan temperatur hidrolisa akan meningkatkan perolehan furfural. Perolehan asam asetat juga meningkat rata-rata hingga 2 kali dari konsentrasi asam asetat awal. Hal ini terjadi karena pemutusan gugus acetyl dari fraksi hemiselulosa pada bagas tebu. Perolehan kondisi optimum yakni pada waktu dan temperatur hidrolisa 2 jam dan 103oC, konsentrasi katalis 9%, dengan konsentrasi furfural 4,10 mg/ml dan konsentrasi asam asetat 2,62 mmol/ml.

  14. Shielding design of highly activated sample storage at reactor TRIGA PUSPATI

    International Nuclear Information System (INIS)

    Naim Syauqi Hamzah; Julia Abdul Karim; Mohamad Hairie Rabir; Muhd Husamuddin Abdul Khalil; Mohd Amin Sharifuldin Salleh

    2010-01-01

    Radiation protection has always been one of the most important things considered in Reaktor Triga PUSPATI (RTP) management. Currently, demands on sample activation were increased from variety of applicant in different research field area. Radiological hazard may occur if the samples evaluation done were misjudge or miscalculated. At present, there is no appropriate storage for highly activated samples. For that purpose, special irradiated samples storage box should be provided in order to segregate highly activated samples that produce high dose level and typical activated samples that produce lower dose level (1 - 2 mR/ hr). In this study, thickness required by common shielding material such as lead and concrete to reduce highly activated radiotracer sample (potassium bromide) with initial exposure dose of 5 R/ hr to background level (0.05 mR/ hr) were determined. Analyses were done using several methods including conventional shielding equation, half value layer calculation and Micro shield computer code. Design of new irradiated samples storage box for RTP that capable to contain high level gamma radioactivity were then proposed. (author)

  15. Selective separation of anaerobic sludge by means of hydrocyclones; Selektive Abtrennung von Anaerobschlamm mit Hydrozyklonen

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, M.; Bohnet, M. [Technische Univ. Braunschweig (Germany). Inst. fuer Verfahrens- und Kerntechnik

    1999-07-01

    In anaerobic waste water cleaning, biomass concentration constitutes a central problem because of long generating times and low biomass sinking speeds. In order to decouple hydraulic retention time from biomass retention time, biomass must be fed back into the reactor. The fact that separation by means of common gravitational separators such as sedimentation tanks and baffle plate thickeners is unspecific results in the enrichment in the reactor of inorganic solids, whose presence is corollary to the anaerobic sludge process. Hence, industry has a great interest in separating anaerobic sludge into organic and inorganic constituents as a means of safeguarding high operating stability and degradation efficiency of anaerobic reactors. Hydrocyclones, permitting selective separation, are an obvious approach. (orig.) [German] Bei der anaeroben Abwasserreinigung ist die Biomassekonzentrierung aufgrund langer Generationszeiten und geringer Sinkgeschwindigkeiten der Biomasse ein zentrales Problem. Zur Entkopplung der hydraulischen Verweilzeit von der Verweilzeit der Biomasse ist eine Rueckfuehrung der Biomasse erforderlich. Da bisher eingesetzte Schwerkraftabscheider, wie Absetzbecken und Lamellenklaerer, unspezifisch trennen, kommt es zu einer Anreicherung anorganischer Feststoffe im Reaktor, die sich prozessbedingt im Anaerobschlamm befinden. So hat die Industrie ein grosses Interesse an einer Auftrennung des Anaerobschlamms in organische und anorganische Bestandteile, um eine hohe Betriebsstabilitaet und Abbauleistung der Anaerobreaktoren zu gewaehrleisten. Hierzu bieten sich Hydrozyklone an, weil mit ihnen eine selektive Trennung moeglich ist. (orig.)

  16. Radioactivity of spent TRIGA fuel

    Energy Technology Data Exchange (ETDEWEB)

    Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P. [Reactor Department, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2015-04-29

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  17. The evolution of the break preclusion concept for nuclear power plants in Germany

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, H. [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany)

    1997-04-01

    In the updating of the Guidelines for PWR`s of the {open_quotes}Reaktor-Sicherheitskommission{close_quotes} (RSK) in 1981 the requirements on the design have been changed with respect to the postulated leaks and breaks in the primary pressure boundary. The major change was a revision in the requirements for pipe whip protection. As a logical consequence of the {open_quotes}concept of basic safety{close_quotes} a guillotine type break or any other break type resulting in a large opening is not postulated any longer for the calculation of reaction and jet forces. As an upper limit for a leak an area of 0, 1 A (A = open cross section of the pipe) is postulated. This decision was based on a general assessment of the present PWR system design in Germany. Since then a number of piping systems have been requalified in the older nuclear power plants to comply with the break preclusion concept. Also a number of extensions of the concept have been developed to cover also leak-assumptions for branch pipes. Furthermore due considerations have been given to other aspects which could contribute to a leak development in the primary circuit, like vessel penetrations, manhole covers, flanges, etc. Now the break preclusion concept originally applied to the main piping has been developed into an integrated concept for the whole pressure boundary within the containment and will be applied also in the periodic safety review of present nuclear power plants.

  18. CRITICALITY ANALYSIS OF URANIUM STORAGE FACILITY WITH FORMATION RACKS

    Directory of Open Access Journals (Sweden)

    Sri Kuntjoro

    2017-03-01

    ANALISIS KRITIKALITAS DI FASILITAS PENYIMPANAN BAHAN URANIUM DENGAN FORMASI PENGATURAN RAK. Bahan uranium dibutuhkan untuk produksi bahan bakar reaktor penelitian dan radioisotop. Bahan uranium sebelum digunakan terlebih dahulu disimpan pada fasilitas penyimpanan. Salah satu prasyarat fasilitas penyimpanan bahan uranium adalah fasilitas tersebut harus dalam kondisi sub-kritis. Bila kondisi kritis terjadi mengakibatkan proses fissi pada bahan uranium tidak terkendali, sehingga akan menimbulkan suhu yang sangat tinggi. Tujuan dari penelitian ini adalah untuk menganalisa kondisi kritikalitas dari fasilitas penyimpanan bahan uranium yang berada di PT. INUKI (Persero untuk menjamin fasilitas tersebut dalam kondisi sub-kritis. Analisis kritikalitas dilakukan menggunakan program MCNP-5 untuk mengetahui tingkat kritikalitas dari tiga fasilitas penyimpanan bahan uranium untuk kondisi awal dan kondisi setelah ditambahkan rak penyimpanan. Untuk fasilitas penyimpanan 1 dan 2 dibuat tiga skenario pengaturan container pada rak penyimpanan, sedangkan pada fasilitas penyimpanan 3 dilakukan 1 skenario.  Hasil ini menunjukkan seluruh fasilitas penyimpanan pada kondisi awal dan setelah ditambah rak penyimpanan dalam kondisi sub-kritis (k-eff<1. Hasil tersebut selanjutnya dipergunakan sebagai dasar untuk menyusun manejemen pengelolaan bahan uranium. Selain itu juga digunakan sebagai dasar untuk pembuatan ijin dari badan pengawas (BAPETEN. Kata Kunci : kritikalitas, fasilitas penyimpanan berbahan uranium,  k-eff

  19. Radioactivity of spent TRIGA fuel

    International Nuclear Information System (INIS)

    Usang, M. D.; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-01-01

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive

  20. Upgrade of Control and Protection System of the Ignalina Nuclear Power Plant Units 1 and 2

    International Nuclear Information System (INIS)

    Wright, Ronald E.; Fletcher, Norman; Sidnev, Victor E.; Bickel, John H.; Vianello, Aldo; Pearsall, Raymond D.

    2003-01-01

    The Ignalina nuclear power plant (NPP) Units 1 and 2 are Soviet-designed, RBMK (Reaktor Bolshoi Moschnosti Kipyashchiy), channelized, large power-type reactors. The original-design electrical capacity for each unit was 1500 MW. Unit 1 began operating in 1983, and Unit 2 was started up in 1987. In 1994, the government of Lithuania agreed to accept grant support for the Ignalina NPP Safety Improvement Program with funding supplied by the Nuclear Safety Account of the European Bank for Reconstruction and Development (EBRD). As conditions for receiving this funding, the Ignalina NPP agreed to prepare a comprehensive safety analysis report that would undergo independent peer review after it was issued. The EBRD Safety Panel oversaw preparation and review of the report. In 1996, the safety analysis report for Unit 1 was completed and delivered to the EBRD. Part of the analyses covered anticipated transients without scram (ATWS). The analysis showed that some ATWS scenarios could lead to unacceptable consequences in <1 min. The EBRD Safety Panel recommended to the government of Lithuania that the Ignalina NPP develop and implement a program of compensatory measures for the control and protection system before the unit would be allowed to return to operation following its 1998 maintenance outage. A compensatory control and protection system that would mitigate the unacceptable consequences was designed, procured, manufactured, tested, and installed. The project was funded by U.S. Department of Energy

  1. Methods for measuring nuclear properties of materials, Safety coefficient method and measurement of effective absorption coefficient of graphite by safety coefficient method; Razvijanje metoda merenja nuklearnih karakteristika materijala, Razrada metode koeficijenta opasnosti i merenje efektivnog apsorpcionog preseka grafita metodom koeficijenta opasnosti

    Energy Technology Data Exchange (ETDEWEB)

    Maglic, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-11-15

    Reactivity of a reactor depends on production, absorption and leaking of neutrons. Change of absorption causes reactivity change, and this fact is used for determining the neutron absorption cross section for the sample inserted in the reactor core. Method for determining the absorption cross section based on reactivity change is called method of safety coefficient. Measurements of neutron absorption cross section for graphite was done in the RA reactor vertical experimental channel VK-5. taking into account the results obtained for five types of graphite this method is considered to be reliable for use. Comparison of nuclear properties of different types of graphite was done as well. Reaktivnost reaktora zavisi od proizvodnje neutrona, apsorpcije i isticanja neutrona. Promena apsorpcije izaziva promene reaktivnosti reaktora pa se ova osobina koristi za odedjivanje neutronskog apsorpcionog preseka uzorka koji se unosi u reaktor. Metoda merenja apsorpcionog preseka na bazi promene reaktivnosti nazvana je metodom koeficijenta opasnosti. Merenje apsorpcionog preseka grafita uradjeno je na reaktoru RA u vertikalnom eksperimentalnom kanalu VK-5. S obzirom na rezultate koji su dobijeni za pet vrsta grafita moze se smatrati da je opravdano koriscenje ove metode. Izvrseno je i poredjenje nuklearnih osobina pomenutih tipova grafita.

  2. Impact Of Secondary-Primary Pumps Operating Sequence On The Electrical Power Supply System

    International Nuclear Information System (INIS)

    Suwoto; Rusdiyanto; Kiswanto

    2001-01-01

    The operating procedure of the reactor cooling system has decided that the primary cooling pump should be operated before secondary cooling pump as known primary-secondary pumps operating sequence. This decision is based on consideration that starting current of the primary pump is higher than secondary pump. Therefore, the primary-secondary pumps operating sequence can avoid the power supply system failure. However, this operating procedure has to take a consequence that in case of primary pump failure, the shutdown time period of the reaktor to be longer caused to re operate the primary pump has required that the running secondary pump should be shutted off. To solve this problem, an impact analysis of the secondary-primary pumps operating sequence on the electric power supply system was carried out to identify the revision possibility of the cooling pump operating procedure. The analysis by discussion of the measuring results of the secondary and primary pump starting current related to another electrical loads has been measured. From discussion it can be concluded that secondary-primary pumps operating sequence has no impact to failure in electric power supply system

  3. Pengaruh penambahan Ca(OH2 pada Proses Pirolisis terhadap Hasil Gasifikasi Batubara Bituminus dengan medium Gas CO2

    Directory of Open Access Journals (Sweden)

    Saripah Sobah

    2014-06-01

    Full Text Available Pemanfaatan batubara melalui proses gasifikasi perlu dikembangkan lebih lanjut karena proses ini dapat dijadikan alternatif untuk menggantikan peranan  gas alam sebagai sumber gas sintesis. Di samping itu, proses ini dapat mengurangi pencemaran lingkungan karena teknologi gasifikasi merupakan teknologi yang bersih dan dapat mengurangi jumlah gas CO2 yang dibuang ke lingkungan. Penelitian ini bertujuan untuk mengetahui pengaruh penambahan Ca(OH2 pada proses pirolisis terhadap hasil gasifikasi arang batubara bituminus dengan medium gas CO2. Reaksi karbon dari arang batubara dengan gas CO2 pada proses gasifikasi merupakan reaksi endotermis dan berlangsung sangat lambat pada suhu di bawah 1000oC sehingga digunakan Ca(OH2 sebagai katalisator. Proses gasifikasi batubara dijalankan dalam reaktor fixed bed. Hasil penelitian menunjukkan bahwa gasifikasi arang batubara dengan penambahan Ca(OH2 pada proses pirolisis memberikan pengaruh terhadap komposisi gas hasil yaitu berkurangnya kadar gas CO2 dan menyebabkan berkurangnya kadar belerang pada arang hasil pirolisis dan gasifikasi. Proses ini juga dapat mengurangi kadar gas CO2 sebesar 63,17% dan untuk  gasifikasi tanpa Ca(OH2 , CO2 dapat dikurangi kadarnya sampai 35,2%.

  4. Pirolisis Campuran Sampah Plastik Polistirena Dengan Sampah Plastik Berlapisan Aluminium Foil (Multilayer

    Directory of Open Access Journals (Sweden)

    Yebi Yuriandala

    2016-04-01

    Full Text Available Sampah plastik yang dulunya merupakan masalah lingkungan, saat ini dapat diubah menjadi bahan bakar alternatif dengan menggunakan proses daur ulang yang memanfaatkan energi panas yaitu pirolisis. Analisis yang digunakan dalam penelitian ini adalah analisis terhadap minyak (liquid yang dihasilkan dari pirolisis sampah plastik Polistitren (PS, plastik berlapisan aluminium foil (kemasan/ multilayer (AL dan campuran plastik tersebut. Penelitian ini dilakukan untuk mendapatkan kuantitas produk dan senyawa kimia yang dihasilkan dari pirolisis sampah plastic PS, kemasan dan campuran keduanya. Penelitian dilakukan dengan menempatkan 50 gram PS (PS, 50 gram plastik berlapisan aluminium foil (AL (multi layer,dan PS dengan campuran 10%, 20%, 30%, 40% AL didalam reaktor pirolisis yang terbuat dari stainless steel berbentuk silinder dengan volume 0,96 m3 dengan temperatur akhir 450oC. hasil penelitian menunjukkan bahwa semakin banyak penambahan Plastik berlapisan aluminium foil maka semakin cepat naiknya temperatur mencapai titik optimum yang ditetapkan (450oC. Sedangkan senyawa kimia yang dihasilkan pada pirolisis yang mengandung PS sebagian besar berupa senyawa aromatic, sedangkan pada pirolisis AL sebagian besar berupa senyawa olefin.

  5. Safety report for the Independent CO{sub 2} loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Vol. VI; Album VI: Izvestaj o sigurnosti za nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u VEK reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-07-01

    First part of the safety report for the Independent CO{sub 2} loop for cooling the samples irradiated in the RA reactor vertical experimental channels contains descriptions of the independent CO{sub 2} loop, system for regulation, measurement and control od the loop parameters, description of the dosimetry system, and the plan for testing the experimental device before start-up. Second part of this analysis describes the influence of of the experimental device on the reactor operation under steady state conditions as follows: influence of the head of the independent coolant loop on the reactivity of the reactor core and influence on the reactor temperature coefficient. Third part of the report includes the analysis of possible accidents during operation of the independent CO{sub 2} coolant loop in the reactor. Izvestaj o sigurnosti za nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u VEK reaktora RA sadrzi u prvom delu: opis nezavisnog kola CO{sub 2}, sistema za regulaciju, merenje i kontrolu parametara nezavisnog kola, sistema dozimetrijske kontrole i plan ispitivanja eksperimentalnih uredjaja pre pustanja u rad. Drugi deo ove analize obuhvata uticaj eksperimentalnog uredjaja na reaktor u normalnom rezimu rada i to: uticaj glave petlje nezavisnog kola CO{sub 2} za hladjenje uzoraka na reaktivnost reaktora i uticaj uredjaja na temperaturni koeficijent reaktora. Treci deo sadrzi analizu mogucih akcidenata u toku rada nezavisnog kola CO{sub 2} u reaktoru.

  6. Fluidized-bed gasification of biomass: Conversion of fine carabon particles in the freeboard; Biomassevergasung in der Wirbelschicht: Umsatz von feinen Kohlenstoffpartikeln im Freeboard

    Energy Technology Data Exchange (ETDEWEB)

    Miccio, F [Ist. Ricerche sulla Combustione-CNR, Napoli (Italy); Moersch, O; Spliethoff, H; Hein, K R.G. [Stuttgart Univ. (Germany). Inst. fuer Verfahrenstechnik und Dampfkesselwesen

    1998-09-01

    The conversion of carbon particles in gasification processes was investigated in a fluidized-bed reactor of the Institute of Chemical Engineering and Steam Boiler Technology of Stuttgart University. This reactor is heated electrically to process temperature, and freeboard coal particles can be sampled using an isokinetic probe. The fuel used in the experiments consisted of beech wood chips. The temperature and air rating, i.e. the main parameters of the process, were varied in order to investigate their influence on product gas quality and carbon conversion. The conversion rate is influenced to a significant extent by grain disintegration and discharge of carbon particles. In gasification conditions, a further conversion process takes place in the freeboard. (orig.) [Deutsch] In dieser Arbeit wird die Umsetzung von Kohlenstoffpartikeln unter Vergasungsbedingungen untersucht. Die Versuche wurden an einem Wirbelschichtreaktor des Instituts fuer Verfahrenstechnik und Dampfkesselwesen der Universitaet Stuttgart durchgefuehrt. Dieser Reaktor wird elektrisch auf Prozesstemperatur beheizt. Mit Hilfe einer isokinetischen Sonde koennen Proben von Kohlenstoffpartikeln im Freeboard genommen werden. Als Brennstoff wurden zerkleinerte Buchenholz-Hackschnitzel eingesetzt. Variiert wurden als Hauptparameter des Prozesses Temperatur und Luftzahl. Untersucht wurde der Einfluss dieser Parameter auf die Qualitaet des Produktgases und die Umsetzung des Kohlenstoffes. Kornzersetzungs- und Austragsvorgaenge von Kohlenstoffpartikeln spielen eine wichtige Rolle fuer den Kohlenstoffumsatz. Unter Vergasungsbedingungen findet im Freeboard eine weitere Umsetzung der Partikel statt. (orig.)

  7. Conceptual Design of On-line Based Licensing Review and Assesment System of Nuclear Installations and Nuclear Materials ('PRIBEN')

    International Nuclear Information System (INIS)

    Melani, Ai; Chang, Soon Heung

    2008-01-01

    At the present Indonesia has no nuclear power plant in operation yet, although it is expected that the first nuclear power plant will be operated and commercially available in around the year of 2016 to 2017 in Muria Peninsula. There are only three research reactors, one nuclear fuel fabrication plant for research reactors, and one experimental fuel fabrication plant for nuclear power, one isotope production facility and some other research facilities. All the facility is under Nuclear Energy Regulatory Agency (BAPETEN) controlling through regulation, licensing and inspection. The organizations operation submits licensing application to BAPETEN before utilizing the facility. According to the regulation before BAPETEN give license they perform review and assessment for the utility application. Based on the review and assessment result, BAPETEN may stipulate, reject, delay or terminate the license. In anticipation of expansion of the nuclear program in Indonesia, BAPETEN should have an integrated and updated system for review and asses the licensing application. For this reason, an expert system for the review and asses the licensing application, so-called PRIBEN (Perizinan Reaktor, Instalasi dan Bahan Nuklir/Licensing of Reactor, Nuclear Installations and Nuclear Materials), is developed which runs on the online-based reality environment

  8. Heavy water pumps; Pumpe D{sub 2}O

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V; Nikolic, M

    1963-12-15

    Continuous increase of radiation intensity was observed on all the elements in the heavy water system during first three years of RA reactor operation. The analysis of heavy water has shown the existence of radioactive cobalt. It was found that cobalt comes from stellite, cobalt based alloy which was used for coating of the heavy water pump discs in order to increase resistance to wearing. Cobalt was removed from the surfaces due to friction, and transferred by heavy water into the reactor where it has been irradiated for 29 876 MWh up to 8-15 Ci/g. Radioactive cobalt contaminated all the surfaces of aluminium and stainless steel parts. This report includes detailed description of heavy water pumps repair, exchange of stellite coated parts, decontamination of the heavy water system, distillation of heavy water. [Serbo-Croat] U toku prve tri godine eksploatacije reaktora RA uocen je neprekidni porast intenziteta zracenja na svim elementima u teskovodnom sistemu. Analizom teske vode utvrdjeno je postojanje radioaktivnog kobalta. Ustanovljeno je da kobalt potice od stelita, legure na bazi kobalta kojim su presvuceni rukavci vratila teskovodnih pumpi radi otpornosi na habanje. Kobalt je trenjem skidan sa povrsina, u toku rada prenosen je teskom vodom u reaktor i ozracivan u toku 29 876 MWh do specificne aktivnosti 8-15 Ci/g. Radioaktivni kobalt je kontaminirao sve povrsine od aluminijuma i nerdjajuceg celika. Ovaj izvestaj sadrzi detaljan opis remonta pumpi, zamene delova teskovodnih pumpi novim delovima bez stelitnog sloja, dekontaminacije teskovodnog sistema, destilacije teske vode.

  9. Opinion about difficulties of RA reactor operation under conditions of high activity of the heavy water system - Annex 2; Prilog 2 - Misljenje o teskocama eksploatacije reaktora RA u uslovima visoke aktivnosti teskovodnog sistema

    Energy Technology Data Exchange (ETDEWEB)

    Nikolic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    It was concluded that reactor the reactor operation is very dangerous for the reactor installation as well as safety of the staff under conditions of heavy water increased activity. Two fundamental arguments in favour of this conclusion are: insufficient possibility of reactor components inspection during maintenance and operation in the future period; difficulties in prevention of accidents that could occur is equally dangerous for the reactor facility and the environment. Cleaning and decontamination of the complete heavy water system is needed before the reactor operation starts in order to avoid possible failures or accidental events. [Serbo-Croat] Zakljuceno je da je eksploatacija reakora u uslovima postojece aktivnosti teskovodnog sistema veoma opasna po sam reaktor i po personal. Dva osnovna razloga u prilog ovog zakljucka su: nedovoljna mogucnost kontrole ispravnosti svih elemenata reaktora u toku remonta i ekspolatacije u predstojecem periodu; teskoca borbe sa udesima, koji bi se eventualno dogodili podjednako je opasna po instalaciju i okolinu. Pre no sto se nastavi sa daljom ekspolatacijom reaktora potrebno ciscenje, dekontaminacija sistema teske vode kako bi se izbegla moguca ostecenja ili akcidentalne situacije.

  10. The physics design of EBR-II; Physique du reacteur EBR-II; Fizicheskij raschet ehksperimental'nogo reaktora - razmnozhitelya EVR-II; Aspectos fisicos del reactor EBR-II

    Energy Technology Data Exchange (ETDEWEB)

    Loewenstein, W. B. [Argonne National Laboratory, Argonne, IL (United States)

    1962-03-15

    hexagonal del reactor los datos obtenidos en geometrias simples idealizadas, analiticas o experimentales. Se compara el rendimiento nuclear, incluso el de reproduccion, del reactor real con el del modelo teorico y se describen las variaciones a largo plazo de la reactividad y de la generacion de energia en la envoltura fertil, refiriendolas a los ciclos propuestos para el combustible y la envoltura fertil. La memoria formula consideraciones sobre la seguridad estudiando en particular la introduccion de indices de reactividad normales y anormales y la consecuencia de supuestos efectos de reactividad, que se basan en el comportamiento fisico de la aleacion combustible y de la estructura del reactor, asi como en la extrapolacion al sistema del EBR-II de los experimentos realizados con el conjunto TREAT. Por ultimo, examina el problema de la fusion del cuerpo del reactor EBR-II. (author) [Russian] Vychisleniya statisticheskogo, dinamicheskogo i dlitel'nogo rezhima reaktivnosti ehksperimental'nog o reaktora-razmnozhitel ya EBR-II dayutsya sovmestno s rezul'tatami i analizom ehksperimentov na reaktore EBR-II po dostizheniyu kritichnosti v sukhom sostoyanii i otdel'nykh ehksperimentov na reaktore ZPR-III. Osoboe vnimanie udelyaetsya problemam fizicheskogo rascheta reaktora, kotorye voznikayut posle opredeleniya printsipial'noj skhemy i do sooruzheniya ili vvoda reaktora v ehkspluatatsiyu. Opisyvayutsya analiz bezopasnosti reaktora i soobrazheniya po otsenke opasnostej, a takzheikh vliyanie na raschet reaktora. V doklade opisyvaetsya sposob ispol'zovaniya modeli EBR-II na osnovanii poluchennykh na reaktore ZPR-III dannykh, a takzhe dannykh sukhoj kritichnosti reaktora EBR-II. EHti ehksperimenty, ikh analiz i teoreticheskie vykladki yavlyayutsya osnovoj dlya opredeleniya fizicheskogo povedeniya reaktornoj sistemy. Bolee podrobno issleduyutsya ogranicheniya, prisushchie primeneniyu ehksperimental'nykh dannykh k rassmatrivaemo j ehnergeticheskoj reaktornoj sisteme. Syuda otnosyatsya

  11. The First Two Years of Operating Experience of the Kahl Nuclear Power Station; Experience acquise pendant les deux premieres annees de fonctionnement de la centrale nucleaire de Kahl; Opyt pervykh dvukh let ehkspluatatsii atomnoj ehlektrostantsii v Kale; Experiencia adquirida en los primeros cuatro anos de funcionamiento de la central nucleoelectrica de Kahl

    Energy Technology Data Exchange (ETDEWEB)

    Bruchner, H. J. [Aeg-Kernenergieanlagen, Frankfurt-am-Main (Germany); Weckesser, A. [Versuchs-Atomkraftwerk Kahl Gmbh, Kahl (Germany)

    1963-10-15

    , une boucle experimentale destinee a l'etude de la surchauffe nucleaire. (author) [Spanish] La central de Kahl constituye la primera central nucleoelectrica europea de propiedad privada, y funciona en carga desde junio de 1961. Esta equipada con un reactor de agua hi que trabaja en ciclo indirecto por circulacion natural. Su capacidad electrica neta asciende a 15 MW y hasta febrero de 1963 habia producido 140 millones de kWh. La memoria revisara la experiencia adquirida durante su funcionamiento, ante todo con el extenso programa de ensayos sobre el comportamiento transitorio y la exploracion gamma. Presentara datos acerca del resultado que han dado en funcionamiento ciertas partes de la central, tales como el dispositivo de accionamiento de las barras de control, el sistema de purificacion de los gases de escape y la turbina. Una vez terminado el programa de ensayos, la planta se exploto en carga basica durante algun tiempo a fin de reunir datos sobre el rendimiento del combustible en la ptactica. Una vez completada esta fase, se instalara en el reactor de Kahl un circuito experimental de sobrecalentamiento nuclear. (author) [Russian] Atomnaya ehlektrostantsiya v Kale, pervaya v Evrope chastnaya atomnaya ehlektrostantsiya, ehkspluatiruetsya pod nagruzkoj s iyunya 1961 goda. Na ehlektrostantsii ustanovlen reaktor s kipyashchej vodoj, kosvennym tsiklom i estestvennoj tsirkulyatsiej. Chistaya ehlektricheskaya moshchnost' reaktora sostavlyaet 15 mgvt. Do fevralya 1963 goda kolichestvo poluchennoj ehnergii sostavilo 140 mln. kvt.ch. Rassmotren opyt ehkspluatatsii, v chastnosti rasshirennaya programma ispytanij: naprimer,povedenie reaktora pri perekhodnom protsesse i kontrol' gamma-izlucheniya. Budut predstavleny rezul'taty izucheniya ehkspluatatsionnoj kharakteristiki nekotorykh komponentov ustanovki, naprimer sistemy privoda reguliruptsikh sterzhnej, sistemy udaleniya gaza i turbiny. Posle osushchestvleniya ehtoj programmy ispytanij ustanovka v techenie nekotorogo vremeni

  12. STUDY ON DISCHARGE HEAT UTILIZATION OF 250 MWe PCMSR TURBINE SYSTEM FOR DESALINATION USING MODIFIED MED

    Directory of Open Access Journals (Sweden)

    Andang Widiharto

    2015-03-01

    Full Text Available PCMSR (Passive Compact Molten Salt Reactor is one type of Advanced Nuclear Reactors. The PCMSR has benefit charasteristics of very efficient fuel use, high safety charecteristic as well as high thermodinamics efficiency. This is due to its breeding capability, inherently safe characteristic and totally passive safety system. The PCMSR design consists of three module, i.e. reactor module, turbine module and fuel management module. Analysis in performed by parametric calculation of the turbine system to calculate the turbine system efficiency and the hat available for desalination. After that the mass and energi balance of desalination process are calculated to calculate the amount of distillate produced and the amount of feed sea water needed. The turbine module is designed to be operated at maximum temperature cycle of 1373 K (1200 0C and minimum temperature cycle of 333 K (60 0K. The parametric calculation shows that the optimum turbine pressure ratio is 4.3 that gives the conversion efficiency of 56 % for 4 stages turbine and 4 stages compressor and equiped with recuperator. In this optimum condition, the 250 MWe PCMSR turbine system produces 196 MWth of waste heat with the temperature of cooling fluid in the range from 327 K (54 0C to 368 K (92 0C. This waste heat can be utilized for desalination. By using MMED desalination system, this waste heat can be used to produce fresh water (distillate from sea water feed. The amount of the destillate produced is 48663 ton per day by using 15 distillation effects. The performance ratio value is 2.8727 kg/MJ by using 15 distillation effects. Keywords: PCMSR, discharged heat, MMED desalination   PCMSR (Passive Compact Molten Salt Reactor merupakan salah satu tipe dari Reaktor Nuklir Maju. PCMSR memiliki keuntungan berupa penggunaan bahan bakar yang sangat efisisien, sifat keselamatan tinggi dan sekaligus efisiensi termodinamika yang tinggi. Hal ini disebabkan oleh kemampuan pembiakan bahan bakar, sifat

  13. MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR

    Directory of Open Access Journals (Sweden)

    Suwardi Suwardi

    2015-03-01

    Full Text Available MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR*. The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared at the CNFT (Center for Nuclear Fuel Technology then a ramp test will be performed. The present work is part of designing first irradiation experiments in the PRTF (Power Ramp Test Facility of RSG-GAS 30 MW reactor. The thermal mechanic of the pin during irradiation has simulated. The geometry variation of pellet and cladding is modeled by taking into account different phenomena such as thermal expansion, densification, swelling by fission product, thermal creep and radiation growth. The cladding variation is modeled by thermal expansion, thermal and irradiation creeps. The material properties are modeled by MATPRO and standard numerical parameter of TRANSURANUS code. Results of irradiation simulation with 9 kW/m LHR indicates that pellet-clad contacts onset from 0.090 mm initial gaps after 806 d, when pellet radius expansion attain 0.015 mm while inner cladding creep-down 0.075 mm. A newer computation data show that the maximum measured LHR of n-UO2 pin in the PRTF 12.4 kW/m. The next simulation will be done with a higher LHR, up to ~ 25 kW/m. MODEL SIMULASI VARIASI GEOMETRI DAN STRESS-STRAIN DARI PROTOTIP BAHAN BAKAR PIN BATAN SELAMA UJI IRADIASI DI REAKTOR RSG-GAS. Pusat Teknologi Bahan Bakar Nuklir (PTBBN telah menyiapkan tangkai (pin bahan bakar pendek perdana yang berisi pelet UO2 alam dalam kelongsong paduan zircaloy untuk dilakukan uji iradiasi daya naik. Penelitian ini merupakan bagian dari perancangan percobaan iradiasi pertama di PRTF (Power Ramp Test Fasility yang terpasang di reaktor serbaguna RSG-GAS berdaya 30 MW. Telah dilakukan pemodelan dan simulasi kinerja termal mekanikal pin selama iradiasi. Variasi geometri pelet dan kelongsong selama pengujian dimodelkan dengan memperhatikan fenomena ekspansi termal

  14. Full scale application of the autotrophic denitrification in trickling filters for treatment of rejection water with high ammonia concentrations from sludge dewatering. Final report; Untersuchungen zur autotrophen Stickstoffentfernung aus ammoniumreichem Filtratwasser der Schlammentwaesserung mit grosstechnischer Realisierung in Tropfkoerpern. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Neumueller, B.; Metzger, J.W.; Pinnekamp, J.

    2003-07-01

    strictly. In a semi-technical anammox-reactor which was operated for several years showed, that the process is very robust. Performance loss is only temporary, if the operating conditions aren't fulfilled. (orig.) [German] Bei kommunalen Klaeranlagen mit anaerober Schlammbehandlung faellt bei der Entwaesserung des stabilisierten Schlammes stark ammoniumhaltiges Prozesswasser an. Die Mitbehandlung dieses Teilstroms aus der Schlammbehandlung in der Hauptstrombiologie der Klaeranlage kann zu einer Erhoehung der Stickstoffbelastung von bis zu 20% fuehren. Eine separate Behandlung von Prozesswaessern kann die Stickstoff-Ablaufwerte verbessern und schafft Reserven in der Hauptstrombiologie. Die autotrophe Denitrifikation nach vorangegangener Teilnitritation in Tropfkoerpern ist ein neuartiges Verfahren zur Behandlung hoch ammoniumhaltiger Abwaesser. Hierzu werden in einem ersten Tropfkoerper etwa 60% des Ammoniums zu Nitrit umgewandelt, wobei die Nitratation gehemmt wird. In einem zweiten geschlossenen Tropfkoerper wird das Ammonium unter anoxischen Bedingungen, mit Nitrit als Elektronenakzeptor, zu molekularem Stickstoff oxidiert. Dieses Verfahren wurde erstmals im grosstechnischen Betrieb auf der Klaeranlage Sindelfingen realisiert. Die Untersuchungen zeigten, dass bereits wenige mg/l Ammoniak im ersten Reaktor ausreichten, um die Nitratation zu hemmen. Da die Nitratation nach einem Anstieg und dem darauffolgenden Abfallen der Ammoniakkonzentration zunahm, fand offensichtlich eine Adaption der Biomasse an die hoeheren Ammoniakgehalte statt. Deshalb sollte die Ammoniakkonzentration moeglichst konstant gehalten und groessere Schwankungen ueber eine laengere Zeit vermieden werden. Nur mit der thermischen Abtoetung der Mikroorganismen und dem erneuten Einfahren des Prozesses konnte nach einer Adaption wieder eine vollstaendige Nitratationshemmung bei geringen Ammoniakkonzentrationen gewaehrleistet werden. Durch die aeusserst geringe Wachstumsrate der Anammoxbakterien nimmt

  15. A CONCEPTUAL DESIGN OF NEUTRON COLLIMATOR IN THE THERMAL COLUMN OF KARTINI RESEARCH REACTOR FOR IN VITRO AND IN VIVO TEST OF BORON NEUTRON CAPTURE THERAPY

    Directory of Open Access Journals (Sweden)

    Nina Fauziah

    2015-03-01

    Full Text Available Studies were carried out to design a collimator which results in epithermal neutron beam for IN VITRO and IN VIVO of Boron Neutron Capture Therapy (BNCT at the Kartini research reactor by means of Monte Carlo N-Particle (MCNP codes. Reactor within 100 kW of thermal power was used as the neutron source. The design criteria were based on recommendation from the International Atomic Energy Agency (IAEA. All materials used were varied in size, according to the value of mean free path for each material. MCNP simulations indicated that by using 5 cm thick of Ni as collimator wall, 60 cm thick of Al as moderator, 15 cm thick of 60Ni as filter, 2 cm thick of Bi as γ-ray shielding, 3 cm thick of 6Li2CO3-polyethylene as beam delimiter, with 1 to 5 cm varied aperture size, epithermal neutron beam with maximum flux of 7.65 x 108 n.cm-2.s-1 could be produced. The beam has minimum fast neutron and γ-ray components of, respectively, 1.76 x 10-13 Gy.cm2.n-1 and 1.32 x 10-13 Gy.cm2.n-1, minimum thermal neutron per epithermal neutron ratio of 0.008, and maximum directionality of 0.73. It did not fully pass the IAEA’s criteria, since the epithermal neutron flux was below the recommended value, 1.0 x 109 n.cm-2.s-1. Nonetheless, it was still usable with epithermal neutron flux exceeding 5.0 x 108 n.cm-2.s-1. When it was assumed that the graphite inside the thermal column was not discharged but only the part which was going to be replaced by the collimator, the performance of the collimator became better within the positive effect from the surrounding graphite that the beam resulted passed all criteria with epithermal neutron flux up to 1.68 x 109 n.cm-2.s-1. Keywords: design, collimator, epithermal neutron beam, BNCT, MCNP, criteria   Telah dilakukan penelitian tentang desain kolimator yang menghasilkan radiasi netron epitermal untuk uji in vitro dan in vivo pada Boron Neutron Capture Therapy (BNCT di Reaktor Riset Kartini dengan menggunakan program Monte

  16. Data about operation and utilization of the RA in 1976 - Report, Annex I; Prilog I - Podaci o radu i iskoriscenosti reaktora RA u 1976. godini - Izvestaj -

    Energy Technology Data Exchange (ETDEWEB)

    Stanic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-12-15

    During 1976 there were significant discrepancies in RA reactor operation in comparison with the operating plan made at the end of 1975. Some discrepancies were planned but not all could be foreseen due to the following reasons: no possibility to estimate the precise schedule of aluminium tubes for fuel channels delivery from the; USSR; no means to estimate precise time for repairs that were never performed or were not done during previous 6 years; inability to estimate the time needed for experiments within the program of transfer to highly enriched fuel (which was successfully finished) since some of these experiments were never done before. Some repairs were not planned since unpredicted failures occurred. Six fuel exchanges were planned, but they were somewhat delayed. Detailed data about the planned and achieved reactor operation are shown in this annex as well as interruptions. There were 12 shorter interruptions and safety shutdowns which was in total for 64 hours. Reactor was operated at nominal power of 6.5 MW. There have been no significant accidents, but two cases could be characterised as accidental according to their occurrence not according to the effects. The first was break of the tube next to the heavy water pump, and the second was contamination of the reactor hall and corridor during preparation of the fuel exchange machine (spill of heavy water). This annex covers the data concerned with utilization of experimental facilities and reactor utilization. [Serbo-Croat] Reaktor RA je u toku 1976. godine radio sa znatnijim odstupanjima od plana rada za 1976. godinu, donetom krajem 1975. godine. Medjutim, ni ova procena mogucih odstupanja nije mogla biti korektna iz sledecih, razloga: nemogucnost da se predvidi tacan rok isporuke cevi od specijalne aluminijumske legure iz SSSR za izradu tehnoloskih kanala, nemogucnost da se predvidi tacno vreme potrebno za remonte kakvi ili nisu radjeni ranije ili su radjeni pre vise od 6 godina pod povoljnijim uslovima

  17. Report on results of the Experts from the Boris Kidric Institute to the Institute for theoretical and experimental physics in Moscow - Operational Report; Radni izvestaj - Izvestaj o rezultatima posete strucnjaka iz IBK Institutu za teorijsku i eksperimentalnu fiziku u Moskvi

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Petrovic, M; Cupac, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-12-15

    .) upoznavanje eksploatacionih karakteristika reaktora TVRS sa 80% obogacenim gorivom, kao i iskustva u vezi sa koriscenjem tog goriva u reaktoru TVRS u Moskvi; b.) upoznavanje sa elementima sigurnosnih analiza u vezi sa uvodjenjem i kriscenjem tog goriva u reaktoru TVRS kao i c.) razmena gledista o programu prevodjenja reaktora RA na novo 80% obogaceno gorivo. U izvestaju su prikazani rezultati ove posete. Osnovni zakljucci do kojih se doslo jesu: a.) poseta je bila korisna i pored toga sto na neka pitanja nije dobijen odgovor; b.) posle obrade dobijenih informacija, posle sticanja prvih iskustava o koriscenju novog goriva u reaktoru RA kao i posle donosenja definitivnog suda o ravnoteznom rezimu reaktora RA sa novim gorivom, bilo bi veoma korisno da se izvrsi jos jedna poseta Institutu u Moskvi; c.) novo disperziono gorivo je mnogo pouzdanije u eksploataciji od starog 2% obogacenog metalnog goriva (kakvo je sada u reaktoru RA); d.) nije bilo primedbi sa aspekta sigurnosti na nas program prevodjenja (prelazni rezim) reaktora RA na novo gorivo; e.) zatrazice se dopunske informacije o maksimalnom dopustivom stepenu izgaranja 80% obogacenog disperzionog goriva i po potrebi izvrsice se blagovremena revizija predvidjenog optimalnog ravnoteznog rezima izgaranja za reaktor RA (baziranog na maksimalnom izgaranju goriva koje omogucuju parametri reaktora RA); f.) metodologija termickog proracuna gorivnih kanala, koja se primenjuje na reaktor RA ocenjena je kao ekvivalentna postupku koji se primenjuje za proracun kanala reaktora TVRS u Moskvi (author)

  18. Pengaruh Pretreatment Secara Alkalisasi-Resistive Heating terhadap Kandungan Lignoselulosa Jerami Padi

    Directory of Open Access Journals (Sweden)

    Dewi Maya Maharani

    2017-09-01

    pretreatment yang berfungsi untuk mendegradasi ikatan lignin, meningkatkan luas permukaan biomassa dan dekristalisasi selulosa. Tujuan dari penelitian ini adalah mengetahui pengaruh alkalisasi resistive heating pada proses pretreatment jerami padi sebelum dikonversi lebih lanjut menjadi bioetanol dan mengetahui pengaruh suhu pemanasan serta konsentrasi NaOH selama pretreatment terhadap perubahan kandungan lignin, selulosa dan hemiselulosa. Sebelum dilakukan penelitian dilakukan perancangan reaktor resistive heating. Jerami padi ukuran 100 mesh dilarutkan pada larutan NaOH dengan variasi konsentrasi 0,03 M, 0,05 M, dan 0,07 M, selanjutnya dipanaskan pada reaktor resistive heating dengan variasi suhu pemanasan 75 oC, 85 oC, dan 99 oC. Selulosa merupakan senyawa yang akan dikonversi lebih lanjut menjadi glukosa. Sehingga pada penelitian ini dipilih kondisi optimum berdasarkan peningkatan selulosa tertinggi hingga 8,88% serta penurunan lignin dan hemiselulosa sebesar 1,39% dan 4,33% pada perlakuan suhu pemanasan 75 oC dan konsentrasi NaOH 0,07 M. Alkalisasi resistive heating dapat diterapkan pada pretreatment jerami padi karena dapat mengurangi kandungan lignin dan hemiselulosa serta meningkatkan kandungan selulosa.

  19. Malaysian Nuclear Agency: Annual report 2008

    International Nuclear Information System (INIS)

    2008-01-01

    The establishment of Malaysian Nuclear Agency (Nuclear Malaysia) was mooted from idea of the then Malaysia's Deputy Prime Minister, Tun Dr. Ismail Dato Abdul Rahman, that Malaysia should play a role in the development of nuclear science and technology for peaceful purposes. The Centre for Application of Nuclear Energy (CRANE) was the entity to mark the of Malaysia's nuclear programme, focussing on manpower development for a nuclear power programme to provide an option for energy source, following the worldwide oil crisis of the early 1970s. The Cabinet officially approved the establishment of the Tun Ismail Atomic Research Centre (PUSPATI), under the Ministry of Science, Technology and the environment on 19 September 1972. The era of nuclear research in Malaysia began with the historic event signified by the Reaktor TRIGA PUSPATI reaching its first criticality on 28 June 1982. When PUSPATI was placed under the auspices of the Prime Ministers Department, it assumed the name Nuclear Energy Unit (UTN). The Nuclear Energy Unit was later placed under the Minister of Science, Technology and the Environment. In line with the national development, the institute was name Malaysian Institute for Nuclear Technology Research (MINT) on 10 August 1994. To reflect its vision, mission, objectives and activities in the challenging world, a new identity was established, and was officially named as Malaysian Nuclear Agency (Nuclear Malaysia) on 28 September 2006. Nuclear Malaysia, is strategically located nearby the government administration, centre Putrajaya, and Cyberjaya. These annual report highlights all the activities that have been through by the agency in 2008. All the achievements and triumph were highlights in this annual report. It also contained all the agency planning during 2008 to fulfill the objectives, mission and vision to become main players in nuclear research in Malaysia. Finally, there also highlights some publications contribute by all the researchers from

  20. Malaysian Nuclear Agency: Annual report 2009

    International Nuclear Information System (INIS)

    2009-01-01

    The establishment of Malaysian Nuclear Agency (Nuclear Malaysia) was mooted from idea of the then Malaysia's Deputy Prime Minister, Tun Dr. Ismail Dato Abdul Rahman, that Malaysia should play a role in the development of nuclear science and technology for peaceful purposes. The Centre for Application of Nuclear Energy (CRANE) was the entity to mark the of Malaysia's nuclear programme, focussing on manpower development for a nuclear power programme to provide an option for energy source, following the worldwide oil crisis of the early 1970s. The Cabinet officially approved the establishment of the Tun Ismail Atomic Research Centre (PUSPATI), under the Ministry of Science, Technology and the environment on 19 September 1972. The era of nuclear research in Malaysia began with the historic event signified by the Reaktor TRIGA PUSPATI reaching its first criticality on 28 June 1982. When PUSPATI was placed under the auspices of the Prime Ministers Department, it assumed the name Nuclear Energy Unit (UTN). The Nuclear Energy Unit was later placed under the Minister of Science, Technology and the Environment. In line with the national development, the institute was name Malaysian Institute for Nuclear Technology Research (MINT) on 10 August 1994. To reflect its vision, mission, objectives and activities in the challenging world, a new identity was established, and was officially named as Malaysian Nuclear Agency (Nuclear Malaysia) on 28 September 2006. Nuclear Malaysia, is strategically located nearby the government administration, centre Putrajaya, and Cyberjaya. These annual report highlights all the activities that have been through by the agency in 2009. All the achievements and triumph were highlights in this annual report. It also contained all the agency planning during 2009 to fulfill the objectives, mission and vision to become main players in nuclear research in Malaysia. Finally, there also highlights some publications contribute by all the researchers from

  1. Radiation protection at the RA Reactor in 1995, Part -2, Annex 2, Decontamination and actions, collection of liquid effluents and solid radioactive waste; Deo 2 - Prilog 2 - Dekontaminacija i intervencije, skupljanje tecnih efluenata i cvrstih radioaktivnih otpadnih materijala

    Energy Technology Data Exchange (ETDEWEB)

    Mandic, M; Vukovic, Z; Lazic, S; Plecas, I; Voko, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1995-12-01

    Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [Serbo-Croat] Tokom rada reaktora RA dolazi do stvaranja odredjenih cvrstih otpadnih materijala cija prosecna kolicina zavisi od vremena rada reaktora i aktivnosti koje se tamo obavljaju. Tokom remonta, kada reaktor ne radi kao i pri akcidentalnim situacijama nastaju vece kolicine otpadnih materijala koje zavise od obima i vrste remontnih operacija i obima dekontaminacije kontaminirane radne povrsine i kontaminiranog alata, predmeta, opreme, itd. Nastali otpadni materijali se razvrstavaju i pakuju na mestu nastanka prema odgovarajucim propisima u skladu sa principima zastite od zracenja i aspekta bezbednosti u cilju minimiziranja nepotrebnog ozracivanja ljudstva za preuzimanje, kontrolu, transport, naknadnu obradu RAO i dekontaminaciju. Pri nerutinskim operacijama (dekontaminacija, remont, kontaminiarni otpadni materijal velike zapremine i sl.), strucna sluzba Institita ZASTITA pruza strucne konsultacije i pomaze pri planiranju

  2. Report of the Technology Service - Annex B; Prilog B - Izvestaj o radu tehnoloske sluzbe

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Kozomara-Maic, S; Cupac, S; Raickovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1982-12-15

    During 1982 the Ra reactor was operated at 2 MW power with the initial core configuration with 80% enriched fuel. The reactor power would be limited until the ventilation system is be repaired and the emergency cooling system built. The limited operation regime during 1982 had twofold application: as regular regime for the users according to their demands and for experiments related to the testing operation at 2 MW power level. The following reactor physics activities were affected by this operational regime: methods for regular operation control (determining power density and burnup distribution, reactivity variation, et,); measurement of utilization parameters (spatial distribution of the neutron flux); analysis of origin and consequences of irregularities during operation (control of heavy water and gas composition by gamma spectrometry, control of fission products, control of the heavy water purification system); safety analyses including future new systems (emergency cooling and ventilation) [Serbo-Croat] Reaktor RA je tokom 1982 godine radio sa pocetnom konfiguracijom jezgra sa 80% obogacenim gorivom na snazi od 2 Mw. Snaga je ogranicena dok se ne izvrsi rekonstrukcija sistema ventilacije i ugradnja sistema za udesno hladjenje. Rezim rada na ogranicenoj snazi koriscen je u 1982. godini dvojako: kao normalni radni rezim za potrebe korisnika i za eksperimenta u okviru probnog rada na snazi 2 MW. Ovim rezimom rada uslovljene su delatnosti u okviru reaktorske fizike na sledece oblasti: metodologija kontrole redovnog radnog rezim (odredjivanje raspodele snage i izgaranja goriva, promene reaktivnosti, itd.), merenje eksploatacionih parametara reaktora (prostorna raspodela neutronskog fluks, i sl,), analiza uzroka i posledica pojava neregularnosti u radu fizickim metodama (kontrola sadrzaja teske vode i gasa gama spektometrijom, kontrola fisionih produkata, kontrola sistema za preciscavanje teske vode), sigurnosne analize ukljucujuci buducu izgradnju novih sistema

  3. Irradiation effects on Zr-2.5Nb in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

  4. Anaerobic degradation of tetrachloroethylene; Anaerober Abbau von Tetrachlorethylen

    Energy Technology Data Exchange (ETDEWEB)

    Diekert, G [Stuttgart Univ. (Germany). Inst. fuer Mikrobiologie; Scholz-Muramatsu, H [Stuttgart Univ. (Germany). Inst. fuer Siedlungswasserbau

    1997-12-31

    Dehalospirillum multivorans, a tetrachloroethylene-dechlorinating bacterium, was isolated in activated sludge. This organism is able to grow on a defined medium with hydrogen and tetrachloroethylene (PCE) as its only energy source. The organism was characterised and the physiology of dechlorination was studied. In this process PCE is dechlorinated to cis-1,2-dichloroethene (DCE) via trichloroethene (TCE). A fluidized-bed reactor which reduces PCE to DCE at a high rate (15 nmol/min/mg of protein at 5 {mu}M PCE) was inoculated with the bacterium. Meanwhile a reactor inoculated with D. multivorans and a fully dechlorinating mixed culture has become available which catalyses the complete dechlorination of PCE to ethene at just as high rates. Tetrachloroethene dehalogenase was purified from D. multivorans (unpublished results) and characterised. (orig./SR) [Deutsch] Aus Belebtschlamm wurde ein Tetrachlorethen-dechlorierendes Bakterium, Dehalospirillum multivorans, isoliert. Der Organismus waechst auf definiertem Medium mit Wasserstoff und Tetrachlorethen (PCE) als einziger Energiequelle. Der Organismus wurde charakterisiert und die Physiologie der Dechlorierung wurde untersucht. PCE wird dabei ueber Trichlorethen (TCE) bis zum cis-1,2-Dichlorethen (DCE) dechloriert. Mit diesem Bakterium wurde ein Wirbelschichtreaktor inokuliert, der mit hohen Raten (15 nmol/min/mg Protein bei 5 {mu}M PCE) PCE zu DCE reduziert. Inzwischen steht ein Reaktor zur Verfuegung, der mit D. multivorans und einer voellig dechlorierenden Mischkultur inokuliert wurde und der mit ebenso hohen Raten eine vollstaendige Dechlorierung von PCE bis zum Ethen katalysiert. Aus D. multivorans wurde die Tetrachlorethen-Dehalogenase gereinigt (unveroeffentlichte Ergebnisse) und charakterisiert. (orig./SR)

  5. Operation, maintenance and utilization of the RA reactor, Annual report 1978; Pogon, odrzavanje i eksploatacija reaktora RA, Izvestaj o radu u 1978. godini

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1978-12-15

    It has been planned for 1978 that the RA reactor would be operated for 158 dana at nominal power of 6.5 MW meaning production of 24 648 MWh. The plan was fulfilled since 24 652 MWh was produces. Reactor operation for 158 days is relevant to reactor operation for 200 days in the period before 1975. The reason is increased neutron flux achieved due to improved fuel management and the characteristics of the new 80% enriched fuel. At the end of 1978 the reactor core contained 45% of 80% enriched fuel elements. Increase of neutron flux has shortened the typical time needed for irradiation of the most important samples for isotope production. This significant success in reactor operation is at the same time an obligation for increasing its utilization. Some new trends proposed for increasing reactor utilization capacities were presented at the Conference on utilization of research nuclear reactors in Yugoslavia held in May 1978. [Serbo-Croat] Reaktor RA imao je u planu za 1978. godinu 158 dana rada na nominalnoj snazi od 6.5 MW, sto odgovara radu od 24 648 MWh. Ostvareno je 24 652 MWh sto znaci da je plan ostvaren. Rad reaktora od 158 dana odgovara radu reaktora od 200 dana u periodu pre 1975. godine. Razlog je povecanje neutronskog fluksa zahvaljujuci usavrsenom rukovanju gorivom i karakteristikama novog 80% obogacenog goriva. Krajem 1978. godine 45% jezgra reaktora bilo je popunjeno novim 80% obogacenim gorivom. Povecani neutronski fluks omogucio je skracenje vremena ozracivanja vaznih uzoraka za proizvodnju radioaktivnih izotopa. Ovaj znacajan uspeh je istovremeno obaveza znatno veceg iskoriscenja reaktora RA. Rezultati napora da se postigne vece iskoriscenje reaktora RA prezentirani su na Konferenciji o koriscenju nuklearnih reaktora u Jugoslaviji koja je odrzana u maju 1978.

  6. Scaling effects on H2-deflagration in containment-geometries - BASSIM. Further development and verification. Final report

    International Nuclear Information System (INIS)

    Rastogi, A.K.; Wennerberg, D.; Fischer, K.

    1998-01-01

    A multidimensional mechanistic calculation procedure for simulating H 2 -deflagration in multiroom geometries was developed at Battelle in a previous project. This calculation method was verified against a number of experiments performed in BMC (Battelle Model Containment) and HDR (Heissdampf Reaktor) facilities. It turned out that the above mentioned procedure overpredicted the H 2 -burnrates in experiments in smaller facilities and therefore was unable to predict the important 'scaling influences'. It is the purpose of the present work to develop the above mentioned calculation procedure BASSIM-H 2 (mod 2.3) further in order to predict the scaling influences correctly. In the present work the combustion model was developed further such that the important phenomena e.g. ignition phase, quasi-laminar initial phase, and the turbulent phase of a H 2 premixed flame would be modelled realistically. The model developed has been verified against 16 very different experiments from 9 different facilities. The computed cases varied in volumes from 0.022 m 3 up to 2100 m 3 . These cases have also been computed with the older model verified in [15]. Based on the comparison between the computed results obtained with the new model and the computed results obtained with the old model as well as with the experimental data, the model put forward in this work is evaluated. The present model computes the scaling effects on H 2 -deflagration satisfactorily with the same set of empirical constants. The flame propagation in horizontal as well as vertical (both upwards and downwards) directions can be computed satisfactorily. The influence of flow obstructions and heat loss at walls is considered as well. (orig.) [de

  7. Radiation protection at the RA Reactor in 1989, Part -2, Decontamination, collection of treatment of fluid and solid radioactive waste, Annex 3; Deo 2 - Zastita od zracenja kod reaktora RA u 1989. godini, Dekontaminacija i intervencija, sakupljanje i obrada tecnih i cvrstih radioaktivnih otpadnih materija za potrebe reaktora RA - Prilog 3

    Energy Technology Data Exchange (ETDEWEB)

    Mandic, M; Vukovic, Z; Plecas, I; Knezevic, Lj; Lazic, S; Bacic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1989-12-15

    Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [Serbo-Croat] Tokom rada reaktora RA dolazi do stvaranja odredjenih cvrstih otpadnih materijala cija prosecna kolicina zavisi od vremena rada reaktora i aktivnosti koje se tamo obavljaju. Tokom remonta, kada reaktor ne radi kao i pri akcidentalnim situacijama nastaju vece kolicine otpadnih materijala koje zavise od obima i vrste remontnih operacija i obima dekontaminacije kontaminirane radne povrsine i kontaminiranog alata, predmeta, opreme, itd. Nastali otpadni materijali se razvrstavaju i pakuju na mestu nastanka prema odgovarajucim propisima u skladu sa principima zastite od zracenja i aspekta bezbednosti u cilju minimiziranja nepotrebnog ozracivanja ljudstva za preuzimanje, kontrolu, transport, naknadnu obradu RAO i dekontaminaciju. Pri nerutinskim operacijama (dekontaminacija, remont, kontaminiarni otpadni materijal velike zapremine i sl.), strucna sluzba Institita ZASTITA pruza strucne konsultacije i pomaze pri planiranju

  8. Pembuatan Biodiesel Dari Minyak Kelapa Dengan Katalis NaOH Menggunakan Gelombang Mikro (Microwave Secara Kontinyu

    Directory of Open Access Journals (Sweden)

    Daru Satria Prayanto

    2016-04-01

    Full Text Available Biodiesel merupakan bioenergi atau bahan bakar nabati yang dibuat dari minyak nabati melalui proses transesterifikasi, esterifikasi, atau proses esterifikasi-transesterifikasi. Proses pembuatan biodiesel dapat dilakukan dengan metode pemanasan konvensional maupun dengan metode pemanasan microwave. Dengan radiasi microwave, maka waktu yang dibutuhkan saat proses transesterifikasi lebih singkat dibandingkan dengan konvensional. Disisi lain, minyak kelapa memiliki potensi yang besar untuk digunakan sebagai bahan baku dalam pembuatan biodiesel karena ketersediaannya yang melimpah. Penelitian ini bertujuan untuk membuat biodiesel dari minyak kelapa secara kontinyu melalui proses transesterifikasi metanol dengan menggunakan radiasi microwave dengan katalis NaOH dan mempelajari pengaruh konsentrasi tiap katalis, daya, dan laju umpan yang digunakan terhadap yield, densitas, dan viskositas biodiesel yang dihasilkan. Dalam penelitian ini di gunakan 3 variabel, yaitu laju umpan 0,73; 1,25; 1,72 ml/s, konsentrasi katalis 0,25; 0,5; 1 (% berat variabel daya microwave 100, 264, 400, 600, dan 800 Watt. Rasio umpan ditentukan pada 1:9. Pada tahap persiapan melarutkan metanol dan katalis sesuai dengan variabel hingga tercampur homogen. Selanjutnya tahap transesterifikasi dengan mencampurkan larutan metanol (metanol dan katalis dengan minyak kelapa dengan mol ratio yang telah ditentukan dan mengatur daya microwave untuk memulai proses transesterifikasi, proses berlangsung secara kontinyu menggunakan mix flow reaktor. Selanjutnya pemisahan hasil transesterifikasi dari gliserol, dilanjutkan dengan tahap pencucian dengan aquadest untuk memisahkan impurities dan katalis yang masih tersisa dalam biodiesel kemudian memanaskan pada oven untuk menguapkan kandungan air dalam biodiesel. Selajutnya menganalisisa hasil biodiesel terhadap yield, densitas, dan viskositasnya. Hasil terbaik dari variabel yang digunakan di atas adalah pada katalis NaOH dengan konsentrasi 1

  9. Operation, maintenance and utilization of the RA reactor - Report on operation in 1979; Pogon, odrzavanje i eksploatacija reaktora RA, Izvestaj o radu u 1979. godini

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M

    1979-12-15

    During 1979 the RA reactor was in operation only three months, i.e. only 24% of the planned activity was achieved. The reactor operation was interrupted in March due to problems that could not be solved by the existing equipment. Alkalinity of the heavy water resulting from the existing ammonia ions could not be removed by the existing distillation system. In addition deposition of aluminium oxyhydrate on the fuel elements was increased. This was noticed during routine control when the reactor was shutdown for refueling. The decision of the Director general of the Institute and sanitary Inspector followed, prohibiting further reactor operation. A separate chapter of this report is devoted to the analysis of the difficulties and possible solution of the problem in cooperation with the experts from different laboratories of the Institute. Aged and damaged instruments at the reactor were not exchange due to lack of budget. During the operation there were no accidents. [Serbo-Croat] U toku 1979. godine plan rada ostvaren je sa 24%, odnosno reaktor je radio samo prva tri meseca. U martu je doslo do prekida rada reaktora zbog pojave koja se nije mogla otkloniti postojecom opremom. Alkalnost teske vode, posledica prisustva amonijum jona, ne moze se odstraniti sistemom destilacije. Pored toga usled dotrajalosti opreme povecano je talozenje aluminijum oksihidroksida na gorivnim elementima. Ova pojava uocena je tokom rutinske kontrole prilikom izmene goriva a usledila je obustava rada resenjem Direktora i sanitarnog inspektora. Posebno poglavlje ovog izvestaja posveceno je analizi teskoca u radu i reaktora i resavanju nastalih problema sa gorivom u saradnji sa saradnicima drugih laboratorija Instituta. Dotrajali uredjaji i oprema nisu zamenjeni usled nedostatka sredstava. U toku rada nije bilo akcidenata.

  10. SINTESA GULA DARI SAMPAH ORGANIK DENGAN PROSES HIDROLISIS MENGGUNAKAN KATALIS ASAM

    Directory of Open Access Journals (Sweden)

    Deddy Irawan

    2012-11-01

    Full Text Available SYNTHESIS OF SUGAR FROM ORGANIC WASTES VIA ACID CATALYSTHYDROLYSIS. Hydrolysis process is an important step from every process to produce biofuel withorganic wastes as raw material. Hydrolysis process with chemical uses hydrochloride acid as catalystin which will transform holocellulose to glucose. Raw material of 100 grams is put into hydrolysisreactor with batch system equipped with pressure control and ratio hydrochloride of 1 : 6 w/v. Thevariables studied were temperature of 110-140oC, HCl concentration of 0.5-1%, time of hydrolysis of15-60 minutes. The sugar concentration was taken and then be analyzed by Nelson-Somogy method.The hydrolysis, which was carried out with the temperature of 120oC, time of 30 minutes, HClconcentration of 0.75%, and the pressure of 6 bar, produced sugar reduction of 27.3 mg/mL and yieldof 15.07%. Proses hidrolisis merupakan satu tahap penting dari rangkaian tahapan proses produksi bahan bakarnabati menggunakan bahan baku sampah organik. Proses hidrolisis secara kimiawi menggunakanHCl sebagai katalis akan mengubah holoselulosa yang terdapat pada sampah organik menjadi gula.Gula yang dihasilkan inilah yang dapat difermentasi menjadi bahan bakar nabati. Bahanbaku sebanyak 100 g dimasukkan dalam reaktor hidrolisis sistem batch yang dilengkapi denganpengukur tekanan dan ditambahkan larutan HCl pada perbandingan 1:6 b/v. Hidrolisis dilakukandengan memvariasikan suhu operasi 100-140oC, waktu proses 15-60 menit, serta konsentrasi HCl 0,5-1%. Hidrolisat yang dihasilkan dianalisis kadar gula menggunakan metode Nelson-somogy. Hasilhidrolisis yang dilakukan pada suhu 120oC selama 30 menit serta konsentrasi HCl 0,75% dan tekananterukur 6 bar menghasilkan gula 27,30 mg/mL dan yield gula sebesar 15,07%.

  11. Malaysian Nuclear Agency; Annual report 2013

    International Nuclear Information System (INIS)

    2013-01-01

    The establishment of Malaysian Nuclear Agency (Nuclear Malaysia) was mooted from idea of the then Malaysia's Deputy Prime Minister, Tun Dr. Ismail Dato Abdul Rahman, that Malaysia should play a role in the development of nuclear science and technology for peaceful purposes. The Centre for Application of Nuclear Energy (CRANE) was the entity to mark the of Malaysia's nuclear programme, focussing on manpower development for a nuclear power programme to provide an option for energy source, following the worldwide oil crisis of the early 1970s. The Cabinet officially approved the establishment of the Tun Ismail Atomic Research Centre (PUSPATI), under the Ministry of Science, Technology and the environment on 19 September 1972. The era of nuclear research in Malaysia began with the historic event signified by the Reaktor TRIGA PUSPATI reaching its first criticality on 28 June 1982. When PUSPATI was placed under the auspices of the Prime Ministers Department, it assumed the name Nuclear Energy Unit (UTN). The Nuclear Energy Unit was later placed under the Minister of Science, Technology and the Environment. In line with the national development, the institute was name Malaysian Institute for Nuclear Technology Research (MINT) on 10 August 1994. To reflect its vision, mission, objectives and activities in the challenging world, a new identity was established, and was officially named as Malaysian Nuclear Agency (Nuclear Malaysia) on 28 September 2006. Nuclear Malaysia, is strategically located nearby the government administration, centre Putrajaya, and Cyberjaya. These annual report highlights all the activities that have been through by the agency in 2013. All the achievements and triumph were highlights in this annual report. It also contained all the agency planning during 2013 to fulfill the objectives, mission and vision to become main players in nuclear research in Malaysia. Finally, there also highlights some publications contribute by all the researchers

  12. Degradasi Zat Warna Pada Limbah Cair Industri Tekstil Dengan Metode Fotokatalitik Menggunakan Nanokomposit Tio2 – Zeolit

    Directory of Open Access Journals (Sweden)

    Siti Naimah

    2014-10-01

    Full Text Available Telah dilakukan penelitian degradasi zat warna pada limbah cair industri tekstil menggunakan metode fotokatalitik dengan penambahan nanokomposit TiO2 - zeolit. Tujuan penelitian ini untuk mengetahui efektifitas kemampuan nanokomposit dalam mendegradasi zat warna serta parameter-parameter yang ditetapkan dalam Peraturan Pemerintah Nomor 82 Tahun 2001 tentang pengelolaan kualitas air dan pengendalian pencemaran air. Zeolit alam diaktivasi terlebih dahulu sebelum dikompositkan dengan TiO2. Perbandingan TiO2 : zeolit yang digunakan pada pembuatan nanokomposit adalah 100:0, 20:80, 40:60, 50:50, 60:40, dan 0:100. Percobaan pendahuluan dilakukan dengan menggunakan limbah cair tekstil buatan yang dibuat dari pewarna Synolon yellow S- G6LS (untuk warna kuning dan B/Blue R 150% special (untuk warna biru, sedangkan limbah cair industri tekstil diambil dari salah satu industri di Bogor. Waktu degradasi zat warna dilakukan dalam reaktor fotokatalitik selama 180 menit. Pada perbandingan TiO2 : zeolit 40:60 didapatkan degradasi zat warna tekstil buatan berwarna kuning maksimal adalah 99,9 % dan zat warna tekstil buatan berwarna biru maksimal 99,8%. Analisis warna menggunakan spektrofotometer dan HPLC. Nanokomposit TiO2 : zeolit 40 : 60 merupakan perbandingan optimal sehingga digunakan pada uji coba limbah cair industri tekstil. Degradasi maksimal warna kuning dengan pengolahan fotokatalitik yang ditambahkan nanokomposit pada limbah cair industri tekstil sebesar 98,4%, sedangkan untuk parameter uji zat organik, TSS, TDS, BOD, COD, dan lemak/minyak diperoleh nilai di bawah baku mutu yang dipersyaratkan. 

  13. {sup 137}CS-determination in game meat from some hunting areas in lower Austria; {sup 137}Cs-Bestimmungen im Wildfleisch aus einigen niederoesterreichischen Jagdrevieren

    Energy Technology Data Exchange (ETDEWEB)

    Ayromlou, S. [Inst. fuer Anorganische Chemie der Univ. Wien (Austria); Tataruch, F. [Forschungsinstitut fuer Wildtierkunde und Oekologie der Veterinaermedizinischen Univ., Wien (Austria)

    2001-07-01

    In 1986, the contamination of some regions in Austria by {sup 137}Cs, due to the Chernobyl accident was relatively high. Among other {sup 137}Cs is taken up by people by the consumption of game. In an area of Lower Austria which is relatively heavily contaminated by Chernobyl fallout since 1986 the {sup 137}Cs-contamination of meat of game-animals was measured every year. Clear differences arose in the load of the single game species whose causes just like the temporal changes are discussed. The highest {sup 137}Cs activity concentration was with 5243 Bq/kg measured on a wild boar. With this activity concentration an annual effective dose of only 0,06 mSv can be estimated for an annual average consumption of one kilogram meat of wild boar. (orig.) [German] Druch den Reaktor-Unfall in Tschernobyl wurde Oesterreich gebietsweise relativ stark mit {sup 137}Cs kontaminiert. Unter anderem gelangt {sup 137}Cs durch den Verzehr von Wildfleisch in den Koerper der Menschen. Daher wurden in einem niederoesterreichischen Gebiet, das 1986 durch Fallout verhaeltnismaessig stark kontaminiert worden war seither jaehrlich erlegte Wildtiere auf den {sup 137}Cs-Gehalt ihres Fleisches hin untersucht. Dabei ergaben sich deutliche Unterschiede in der Belastung der einzelnen Wildarten, deren Ursachen ebenso wie die zeitlichen Veraenderungen diskutiert werden. Die hoechste {sup 137}Cs-Aktivitaetskonzentration wurde mit 5243 Bq/kg bei einem Wildschwein gemessen. Mit einen durchschnittlichen Verzehr von 1 kg Wildschweinfleisch pro Jahr kann daraus eine maximale Effektivdosis von nur 0,06 mSv/Jahr abgeschaetzt werden. (orig.)

  14. Anaerobic degradation of tetrachloroethylene; Anaerober Abbau von Tetrachlorethylen

    Energy Technology Data Exchange (ETDEWEB)

    Diekert, G. [Stuttgart Univ. (Germany). Inst. fuer Mikrobiologie; Scholz-Muramatsu, H. [Stuttgart Univ. (Germany). Inst. fuer Siedlungswasserbau

    1996-12-31

    Dehalospirillum multivorans, a tetrachloroethylene-dechlorinating bacterium, was isolated in activated sludge. This organism is able to grow on a defined medium with hydrogen and tetrachloroethylene (PCE) as its only energy source. The organism was characterised and the physiology of dechlorination was studied. In this process PCE is dechlorinated to cis-1,2-dichloroethene (DCE) via trichloroethene (TCE). A fluidized-bed reactor which reduces PCE to DCE at a high rate (15 nmol/min/mg of protein at 5 {mu}M PCE) was inoculated with the bacterium. Meanwhile a reactor inoculated with D. multivorans and a fully dechlorinating mixed culture has become available which catalyses the complete dechlorination of PCE to ethene at just as high rates. Tetrachloroethene dehalogenase was purified from D. multivorans (unpublished results) and characterised. (orig./SR) [Deutsch] Aus Belebtschlamm wurde ein Tetrachlorethen-dechlorierendes Bakterium, Dehalospirillum multivorans, isoliert. Der Organismus waechst auf definiertem Medium mit Wasserstoff und Tetrachlorethen (PCE) als einziger Energiequelle. Der Organismus wurde charakterisiert und die Physiologie der Dechlorierung wurde untersucht. PCE wird dabei ueber Trichlorethen (TCE) bis zum cis-1,2-Dichlorethen (DCE) dechloriert. Mit diesem Bakterium wurde ein Wirbelschichtreaktor inokuliert, der mit hohen Raten (15 nmol/min/mg Protein bei 5 {mu}M PCE) PCE zu DCE reduziert. Inzwischen steht ein Reaktor zur Verfuegung, der mit D. multivorans und einer voellig dechlorierenden Mischkultur inokuliert wurde und der mit ebenso hohen Raten eine vollstaendige Dechlorierung von PCE bis zum Ethen katalysiert. Aus D. multivorans wurde die Tetrachlorethen-Dehalogenase gereinigt (unveroeffentlichte Ergebnisse) und charakterisiert. (orig./SR)

  15. Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters; Fizicka merenja na reaktoru RA u vezi projekta VISA-2 - I deo, Pustanje u rad reaktora RA i merenje fizickih parametara novog jezgra reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-15

    The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included. [Serbo-Croat] Svrha merenja je odredjivanje neutronskog fluksa u reaktoru RA. S obzirom na uvecani broj tehnoloskih kanala of 56 na 68 u vezi projekta VISA-2, bilo je potrebno ponovo dovesti reaktora RA do kriticnosti i izvrsiti merenja karakteristika fluksa neutrona. Posebno je pripremljen 'program pustanja u pogon reaktora RA', koji je sadrzan u ovom dokumentu. Program merenja bio je podeljen na dve faze. Prva faza je merenje fluksa pre podizanju reaktora na nominalnu snagu. Slicna merenja vrsena su i na vecim snagama u drugoj fazi, pod uslovima ravnoteznog zatrovanja reaktora ksenonom, jer se tada pokazuju izvesne promene u odgovarajucim karakteristikama fluksa neutrona. Ovaj izvestaj sadrzi merene vrednosti raspodele fluksa i apsolutne vrednosti termalnih i brzih neutrona kao i kadmijumskih odnosa koji su korisceni za odredjivanje fluksa epitermalnih neutrona. Opisana je kalibracija regulacionih sipki za hladan nezatrovan reaktor.

  16. Adaptive Neural Network Algorithm for Power Control in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Husam Fayiz, Al Masri

    2017-01-01

    The aim of this paper is to design, test and evaluate a prototype of an adaptive neural network algorithm for the power controlling system of a nuclear power plant. The task of power control in nuclear reactors is one of the fundamental tasks in this field. Therefore, researches are constantly conducted to ameliorate the power reactor control process. Currently, in the Department of Automation in the National Research Nuclear University (NRNU) MEPhI, numerous studies are utilizing various methodologies of artificial intelligence (expert systems, neural networks, fuzzy systems and genetic algorithms) to enhance the performance, safety, efficiency and reliability of nuclear power plants. In particular, a study of an adaptive artificial intelligent power regulator in the control systems of nuclear power reactors is being undertaken to enhance performance and to minimize the output error of the Automatic Power Controller (APC) on the grounds of a multifunctional computer analyzer (simulator) of the Water-Water Energetic Reactor known as Vodo-Vodyanoi Energetichesky Reaktor (VVER) in Russian. In this paper, a block diagram of an adaptive reactor power controller was built on the basis of an intelligent control algorithm. When implementing intelligent neural network principles, it is possible to improve the quality and dynamic of any control system in accordance with the principles of adaptive control. It is common knowledge that an adaptive control system permits adjusting the controller’s parameters according to the transitions in the characteristics of the control object or external disturbances. In this project, it is demonstrated that the propitious options for an automatic power controller in nuclear power plants is a control system constructed on intelligent neural network algorithms. (paper)

  17. Operation and maintenance of RA Reactor, Annual report 1977; Pogon i odrzavanje reaktora RA - Izvestaj o radu u 1977. godini

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1977-12-15

    During 1977, the RA Reactor was operated at nominal power of 6.5 MW for 183 days. Total production was 28582 MWh which is 10% higher than planned. Second phase of introducing the 80% enriched fuel was fulfilled according to the plan. This means that the reactor core will be filled with highly enriched fuel in 1978. Refueling was done three time during the past year. After completing the first phase of the fuel exchange which was related mostly to reactor safety, the second phase will be devoted to the more efficient increase of neutron flux . This second phase is of utmost importance because higher neutron flux will provide better and more efficient reactor application from economic point of view. This will justify the application of the new more expensive highly enriched fuel. The budget for reactor operation and maintenance is hardly enough to cover the maintenance of the components and instrumentation. During 1977 there were no accidents related nor incidents related to the instrumentation or related to radiation protection. [Serbo-Croat] Reaktor RA je u toku 1977. godine ostvario rad od 28583 MWh odnosno 183 dana rada na nominalnoj snazi, sto u odnosu na plan rada iznosi 10% vise od planiranog. Druga faza uvodjenja 80% obogacenog goriva izvrsena je prema planu sto znaci da ce se punjenje jezgra reaktora 80% gorivom zavrsiti u toku 1978. godine. Izvrsene su tri izmene goriva. Posle isteka prvog dela prelaznog rezima koji je usmeren na maksimalnu sigurnost reaktora preci ce se na drugi deo usmeren na vece i brze povecanje neutronskog fluksa. Druga faza je od izuzetnog znacaja jer ce omoguciti bolje i znatno ekonomicnije koriscenje reaktora i opravdati upotrebu novog i skupljeg goriva. Sredstva za pogon i odrzavanje reaktora RA su jedva dovoljna za odrzavanje neophodnog nivoa opreme. Akcidenata u toku 1977. godine nije bilo ni sa opremom ni u pogledu zastite od zracenja.

  18. Nuclear Reactor RA Safety Report, Vol. 9, Radiation protection; Izvestaj o sigurnosti nuklearnog reaktora RA, Knjiga 9, Zastita od zracenja

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-11-01

    Instrumentation for Radiation protection existing at the RA reactor is dating mostly from the period 1957-1959 when the reactor has been built. With some minor exception it was produced in USSR. Radiation protection system was constructed based on specific design project, somewhat modified original USSR project which has been indispensable because of some modification of the building design. During the past 27 years no renewal of the instrumentation was done, only maintenance was performed. Instrumentation consists of old electronic devices which caused difficulties and even prevented regular maintenance because of lack of spare parts. Instrumentation for radiation protection at the RA reactor is classified as follows: centralized dosimetry system; stationary dosimetry instrumentation, movable and personal dosimetry systems. Apart from the scheme of dosimetry instrumentation this volume includes description of radiation protection procedures; protection devices; radiation doses and dose limit data; program for environmental radioactivity control; medical control procedures. [Serbo-Croat] Instrumentacija za zastitu od zracenja koja danas postoji na reaktoru RA najvecim delom potice iz perioda 1957-1959 kada je reaktor gradjen. Sa malim izuzetcima instrumentacija je proizvedena u SSSR. Sistem za zastitu izveden je na osnovu posebnog projekta koji predstavlja modifikaciju originalnog projekta koja je bila neophodna usled modifikovanog gradjevinskog projekta. U proteklom periodu od 27 godina instrumentacija nije obnavljana vec je vrseno samo odrzavanje. Instrumentacija je izradjena u danas prevazidjenoh tehnoligiji (elektronske cevi) sto je otezavalo i skoro onemogucavalo normalno odrzavanje. Instrumentacija za zastitu od zracenja na reaktoru RA moze se podeliti na tri dela: centralizovani dozimetrijski sistem; stacionarna dozimetrijska instrumentacija; prenosna i licna dozimetrijska instrumentacija. Pored seme sistema dozimetrijske kontrole ova knjiga sadrzi opis

  19. Malaysian Nuclear Agency: Annual report 2010

    International Nuclear Information System (INIS)

    2010-01-01

    The establishment of Malaysian Nuclear Agency (Nuclear Malaysia) was mooted from idea of the then Malaysia's Deputy Prime Minister, Tun Dr. Ismail Dato Abdul Rahman, that Malaysia should play a role in the development of nuclear science and technology for peaceful purposes. The Centre for Application of Nuclear Energy (CRANE) was the entity to mark the of Malaysia's nuclear programme, focussing on manpower development for a nuclear power programme to provide an option for energy source, following the worldwide oil crisis of the early 1970s. The Cabinet officially approved the establishment of the Tun Ismail Atomic Research Centre (PUSPATI), under the Ministry of Science, Technology and the environment on 19 September 1972. The era of nuclear research in Malaysia began with the historic event signified by the Reaktor TRIGA PUSPATI reaching its first criticality on 28 June 1982. When PUSPATI was placed under the auspices of the Prime Ministers Department, it assumed the name Nuclear Energy Unit (UTN). The Nuclear Energy Unit was later placed under the Minister of Science, Technology and the Environment. In line with the national development, the institute was name Malaysian Institute for Nuclear Technology Research (MINT) on 10 August 1994. To reflect its vision, mission, objectives and activities in the challenging world, a new identity was established, and was officially named as Malaysian Nuclear Agency (Nuclear Malaysia) on 28 September 2006. Nuclear Malaysia, is strategically located nearby the government administration, centre Putrajaya, and Cyberjaya. These annual report highlights all the activities that have been through by the agency in 2010. All the achievements and triumph were highlights in this annual report. It also contained all the agency planning during 2010 to fulfill the objectives, mission and vision to become main players in nuclear research in Malaysia. Finally, there also highlights some publications contribute by all the researchers from

  20. RA reactor kinetic parameters - Progress report; Kineticki parametri reaktora RA - Izvestaj o napredovanju -

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, M; Obradovic, D; Jevtovic, V; Velickovic, Lj [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    The objective of nuclear reactor kinetics study is to analyze the stability of reactor operation in practice. The obtained parameters should define the needed properties of automatic control system relevant for the stability of the designed reactor system. Refining the analytical models is done by using the analysis and interpretation of experimental data. Results of measured the reactor response obtained by using the reactor oscillator ROB-1 are explained by using the space independent model of the zero power reactor, by power reactor model with one feedback circuit, and by a complex model. It was assumed that the perturbations of the system are small and that linearized kinetic equations could be used. Linearized kinetic equation of the reactor system are transformed into the frequency region in order to analyze the measured values directly. The objective of this paper is to measure the RA reactor kinetics parameters, and analyze the stability of reactor operation at power levels high than nominal. Istrazivanja u oblasti kinetike nuklearnih reaktora imaju za cilj da dovedu analizu stabilnosti rada reaktora na nivo 'radne tehnologije'. Dobijeni pararametri treba da specificiraju potrebne karakteristike sistema automatske kontrole za odgovarajucu stabilnost projektovanog reaktorskog sistema. Doterivanjem analitickih modela do takvog nivoa da se zapazeni fenomeni mogu anailitcki predvideti ide preko analize i interpretacije eksperimentalnih podataka. Eksperimentalni rezultati merenja odziva reaktora, izvedeni reaktorskim oscilatorom ROB-1, interpretirani su na osnovu prostorno nezavisnog modela za reaktor nulte snage, modelom reaktora snage sa jednim kolom povratne sprege, kao i kompleksnim modelom. U ovom radu se poslo od toga da su perturbacije parametara sistema male, pa se mogu upotrebiti linearizovane kineticke jednacine. Linearizovane kineticke jednacine reaktorskog sistema transformirane su u frekventno podrucje s ciljem direktne analize mernih rezultata

  1. Investigation of mass transfer phenomena in biofilm systems; Untersuchung von Stoffuebergangsphaenomenen in Biofilmsystemen

    Energy Technology Data Exchange (ETDEWEB)

    Waesche, S.; Hempel, D.C. [Technische Univ. Braunschweig (Germany). Inst. fuer Bioverfahrenstechnik; Horn, H. [Fachhochschule Magdeburg (Germany). Hydro- und Abfallchemie

    1999-07-01

    bestimmt. Die Biofilmdichte (Biotrockenmasse/Biofilmvolumen) wurde gravimetrisch bestimmt. Die maximalen Massestromdichten der Biofilme wurden in Batchversuchen bei Substratueberschuss ermittelt. Mit Hilfe von Sauerstoffmikroelektroden wurden Sauerstoffprofile im Biofilm direkt im Reaktor gemessen. Die Messungen der Sauerstoffprofile wurden bei einer Biofilmdicke von 400-2000 {mu}m durchgefuehrt, bei der der Biofilm noch keine Erosionserscheinungen zeigte. (orig.)

  2. Malaysian Nuclear Agency; Annual report 2014

    International Nuclear Information System (INIS)

    2009-01-01

    The establishment of Malaysian Nuclear Agency (Nuclear Malaysia) was mooted from idea of the then Malaysia's Deputy Prime Minister, Tun Dr. Ismail Dato Abdul Rahman, that Malaysia should play a role in the development of nuclear science and technology for peaceful purposes. The Centre for Application of Nuclear Energy (CRANE) was the entity to mark the of Malaysia's nuclear programme, focussing on manpower development for a nuclear power programme to provide an option for energy source, following the worldwide oil crisis of the early 1970s. The Cabinet officially approved the establishment of the Tun Ismail Atomic Research Centre (PUSPATI), under the Ministry of Science, Technology and the environment on 19 September 1972. The era of nuclear research in Malaysia began with the historic event signified by the Reaktor TRIGA PUSPATI reaching its first criticality on 28 June 1982. When PUSPATI was placed under the auspices of the Prime Ministers Department, it assumed the name Nuclear Energy Unit (UTN). The Nuclear Energy Unit was later placed under the Minister of Science, Technology and the Environment. In line with the national development, the institute was name Malaysian Institute for Nuclear Technology Research (MINT) on 10 August 1994. To reflect its vision, mission, objectives and activities in the challenging world, a new identity was established, and was officially named as Malaysian Nuclear Agency (Nuclear Malaysia) on 28 September 2006. Nuclear Malaysia, is strategically located nearby the government administration, centre Putrajaya, and Cyberjaya. These annual report highlights all the activities that have been through by the agency in 2014. All the achievements and triumph were highlights in this annual report. It also contained all the agency planning during 2014 to fulfill the objectives, mission and vision to become main players in nuclear research in Malaysia. Finally, there also highlights some publications contribute by all the researchers

  3. Preliminary design of RDE feedwater pump impeller

    International Nuclear Information System (INIS)

    Sri Sudadiyo

    2018-01-01

    Nowadays, pumps are being widely used in the thermal power generation including nuclear power plants. Reaktor Daya Experimental (RDE) is a proposed nuclear reactor concept for the type of nuclear power plant in Indonesia. This RDE has thermal power 10 MW th , and uses a feedwater pump within its steam cycle. The performance of feedwater pump depends on size and geometry of impeller model, such as the number of blades and the blade angle. The purpose of this study is to perform a preliminary design on an impeller of feedwater pump for RDE and to simulate its performance characteristics. The Fortran code is used as an aid in data calculation in order to rapidly compute the blade shape of feedwater pump impeller, particularly for a RDE case. The calculations analyses is solved by utilizing empirical correlations, which are related to size and geometry of a pump impeller model, while performance characteristics analysis is done based on velocity triangle diagram. The effect of leakage, pass through the impeller due to the required clearances between the feedwater pump impeller and the volute channel, is also considered. Comparison between the feedwater pump of HTR-10 and of RDE shows similarity in the trend line of curve shape. These characteristics curves will be very useful for the values prediction of performance of a RDE feedwater pump. Preliminary design of feedwater pump provides the size and geometry of impeller blade model with 5-blades, inlet angle 14.5 degrees, exit angle 25 degrees, inside diameter 81.3 mm, exit diameter 275.2 mm, thickness 4.7 mm, and height 14.1 mm. In addition, the optimal values of performance characteristics were obtained when flow capacity was 4.8 kg/s, fluid head was 29.1 m, shaft mechanical power was 2.64 kW, and efficiency was 52 % at rotational speed 1750 rpm. (author)

  4. PIROLISIS LIGNIN DARI LIMBAH INDUSTRI KELAPA SAWIT UNTUK PENGEMBANGAN SURFAKTAN DALAM PROSES ENHANCE OIL RECOVERY (EOR (Pyrolysis of Lignin From Waste of Palm Oil Industries for The Development of Surfactants for Enhance Oil Recovery (EOR

    Directory of Open Access Journals (Sweden)

    Suryo Purwono

    2001-12-01

    Full Text Available ABSTRAK Pirolisis dari lignin yang berasal dari limbah industri kelapa sawit dapat menghasilkan alkohol dan derivatif lainnyd yang dapat digunakan sehagai surfaktan. Prosedur penelitian proses pirolisis ini odalah sebagai berikut: I serabut atau tandan sisa pengolahon kelapa sawit yang sudah dikeringkan dimasukkan kedalam reaktor dengan berat tertentu dan dipanaskan sampai suhu yang diinginkan, 2 produk pirolisis yang keluar dari reoktor kemudian didinginkan sampoi mencapai suhu kamor, 3 hasil cair ditampung didalam gelas ukur dan hasil gasnya ditampung di suatu botol tertentu. Suhu paling baik yang dicapai adalah 4A0 "C untuk lignin yong berasal dari serabut dan 350'C untuk lignin yang berasal dari tandan kelapa sawit. Surfaktan yang dihasilkan sekitar j4 sampai 38% dari produk pirolisis. Pada penelitian ini kecepatan reaksi dianggap order satu. Hasil penelitian menunjukkan bahwa surfakton yang dihasilkan dapat membentuk emulsi dengan minyak menta.h. Hal ini menunjukkon bahwa surfaktan yang dihasilkan dapat digunakan sebagai bahan untuk proses EOR.   ABSTRACT Pyrolysis of lignin from waste of palm oil industries produces alcohol and its derivatives which can be sulfonated to become surfactant. The experimental procedures for the pyrolysis process were as follows: 1 dried palm oil husks at a certain weight were put into the pyrolysis reactor and heated up to a certain temperafure; 2 the product leaving the reactor was cooled down to room temperature; and 3 the liquid product was collected in a flask while the gas product was put into a big bottle. The best temperature obtained for producing liquid product was 400 oC for lignin from palm oil fruit fibers and 350 oC for lignin from palm oil fruit stems. The surfactant developed was in the range between 34 and 38% from the pyrolysis product. In this experiment, the reaction rate was assumed to be in first order. The result showed that the surfactant obtained from the experiment could form emulsion

  5. Malaysian Nuclear Agency; Annual report 2011

    International Nuclear Information System (INIS)

    2008-01-01

    The establishment of Malaysian Nuclear Agency (Nuclear Malaysia) was mooted from idea of the then Malaysia's Deputy Prime Minister, Tun Dr. Ismail Dato Abdul Rahman, that Malaysia should play a role in the development of nuclear science and technology for peaceful purposes. The Centre for Application of Nuclear Energy (CRANE) was the entity to mark the of Malaysia's nuclear programme, focussing on manpower development for a nuclear power programme to provide an option for energy source, following the worldwide oil crisis of the early 1970s. The Cabinet officially approved the establishment of the Tun Ismail Atomic Research Centre (PUSPATI), under the Ministry of Science, Technology and the environment on 19 September 1972. The era of nuclear research in Malaysia began with the historic event signified by the Reaktor TRIGA PUSPATI reaching its first criticality on 28 June 1982. When PUSPATI was placed under the auspices of the Prime Ministers Department, it assumed the name Nuclear Energy Unit (UTN). The Nuclear Energy Unit was later placed under the Minister of Science, Technology and the Environment. In line with the national development, the institute was name Malaysian Institute for Nuclear Technology Research (MINT) on 10 August 1994. To reflect its vision, mission, objectives and activities in the challenging world, a new identity was established, and was officially named as Malaysian Nuclear Agency (Nuclear Malaysia) on 28 September 2006. Nuclear Malaysia, is strategically located nearby the government administration, centre Putrajaya, and Cyberjaya. These annual report highlights all the activities that have been through by the agency in 2011. All the achievements and triumph were highlights in this annual report. It also contained all the agency planning during 2011 to fulfill the objectives, mission and vision to become main players in nuclear research in Malaysia. Finally, there also highlights some publications contribute by all the researchers from

  6. Experimental investigation and mathematical modelling of the combustion of brown coal, refuse and mixed fuels in a circulating fluidized bed combustor; Experimentelle Untersuchung und mathematische Modellierung der Verbrennung von Braunkohle, Abfallstoffen und Mischbrennstoffen in einer zirkulierenden Wirbelschichtfeuerung

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, W; Brunne, T; Hiller, A [Technische Univ. Dresden (Germany). Inst. fuer Energietechnik; Albrecht, J [Lurgi Umwelt GmbH, Frankfurt am Main (Germany); Quang, N [Polytechnic Inst., Danang (Viet Nam)

    1998-09-01

    Extensive experiments on combustion of biological materials and residues in fluidized bed combustors and dust combustors have been carried out at the Department of Power Plant Engineering of Dresden University since the early nineties. Particular interest was taken in mixing brown coal with sewage sludge, sugar pulp and waste wood. The experiments were supplemented by modelling in a research project funded jointly by the BMBF and Messrs. Lurgi since early 1997. A combustion cell model designed by Siegen University is being modified for the new mixed fuels, and preliminary investigations were carried out on a batch reactor while the modelling work was continued. (orig.) [Deutsch] An dem Lehrstuhl fuer Kraftwerkstechnik der TU Dresden werden seit Anfang der 90-iger Jahre umfangreiche experimentelle Untersuchungen zur Verbrennung von Bio- und Reststoffen in Wirbelschicht- und Staubfeuerungen durchgefuehrt. Dabei war vor allem die Zufeuerung dieser Stoffe in Waermeerzeugeranlagen auf Braunkohlenbasis von besonderem Interesse. Experimentell konnte nachgewiesen werden, dass sowohl Biobrennstoffe als auch Abfaelle in zirkulierenden Wirbelschichtfeuerungen umweltschonend zur Waermeerzeugung eingesetzt werden koennen. Als Beispiel wird das an Hand von Braunkohle-Klaerschlammgemischen sowie Bagasse- und Holz-Braunkohlegemischen gezeigt. Neben den experimentellen Untersuchungen bietet die Modellierung der Verbrennungsvorgaenge ein geeignetes Mittel um Voraussagen zu anderen Mischungsanteilen sowie anderen geometrischen Abmessungen machen zu koennen. Seit Anfang 1997 wird dazu ein vom BMBF und der Firma Lurgi gefoerdertes Forschungsvorhaben bearbeitet. Ein von der Universitaet Gesamthochschule Siegen fuer die Braunkohleverbrennung konzipiertes Zellenmodell wird auf die neuen Brennstoffgemische erweitert. Da grundsaetzlich andere Stoffzusammensetzungen vorliegen, wurden an einem Batch-Reaktor Voruntersuchungen zum Pyrolyseverhalten der Brennstoffe durchgefuehrt. Erste

  7. PENGARUH NILAI BAKAR TERHADAP INTEGRITAS KELONGSONG ELEMEN BAKAR TRIGA 2000

    Directory of Open Access Journals (Sweden)

    K.A. Sudjatmi

    2015-04-01

    Full Text Available Bentuk elemen bakar reaktor TRIGA Bandung adalah silinder padat yang merupakan campuran homogen paduan uranium dan zirkonium hidrida. Pada saat reaktor beroperasi, suhu elemen bakar akan bertambah, akibatnya akan menaikan tekanan gas-gas yang ada di dalam kelongsong elemen bakar. Tekanan gas yang timbul dalam kelongsong elemen bakar merupakan penjumlahan tiga komponen tekanan yaitu tekanan akibat udara yang terperangkap antara kelongsong dengan bahan bakar, tekanan oleh gas hasil fisi yang terbentuk dari elemen bakar dan tekanan yang berasal dari pemisahan hidrogen dari paduan zirkonium hidrida. Gas hasil fisi yang terbentuk oleh bahan bakar sebanding dengan besarnya fraksi bakar oleh setiap elemen bakar dalam teras reaktor. Semakin besar fraksi bakar elemen bakar, semakin besar gas gas hasil fisi yang dihasilkannya, akibatnya semakin besar tekanan di dalam kelongsong yang disebabkan oleh gas gas hasil fisi tersebut. Perhitungan jumlah gas-gas hasil fisi dalam kelongsong yang merupakan fungsi dari nilai bakar dilakukan dengan menggunakan program ORIGEN-2. Program ORIGEN-2 adalah kode komputer yang banyak digunakan untuk menghitung hasil fisi, peluruhan dan pengolahan bahan radioaktif. Tampang lintang, presentase timbulnya hasil fisi, data peluruhan, dan data lainnya yang diperlukan disediakan dalam pustaka data selama eksekusi program. Dari hasil perhitungan dapat disimpulkan bahwa tekanan gas yang diakibatkan oleh gas hasil fisi adalah 4,13 10-3 psi dan tekanan gas yang diakibatkan udara yang terjebak di dalam kelongsong adalah 56,6 psi, yang mengakibatkan tegangan pada kelongsong sebesar 2080 psi dan nilai ini jauh lebih kecil dari setengah tegangan luluh bahan kelongsong sebesar 12.000 psi pada temperatur 750 oC atau sekitar 40.000 psi pada temperatur 138 oC. Akhirnya dapat disimpulkan bahwa dilihat dari sisi nilai bakar, maka elemen bakar layak digunakan sampai mencapai nilai bakar maksimum. Kata kunci : TRIGA, nilai bakar, elemen bakar

  8. ANALISIS EFEK KECELAKAAN WATER INGRESS TERHADAP REAKTIVITAS DOPPLER TERAS RGTT200K

    Directory of Open Access Journals (Sweden)

    Zuhair Zuhair

    2015-03-01

    Full Text Available Dalam high temperature reactor, koefisien reaktivitas temperatur yang didesain negatif menjamin reaksi fisi dalam teras tetap berada di bawah kendali dan panas peluruhan tidak akan pernah melelehkan bahan bakar yang menyebabkan terlepasnya zat radioaktif ke lingkungan. Namun masuknya air (water ingress ke dalam teras reaktor akibat pecahnya tabung penukar panas generator uap, yang dikenal sebagai salah satu kecelakaan dasar desain, dapat mengintroduksi reaktivitas positif dengan potensi bahaya lainnya seperti korosi grafit dan kerusakan material struktur reflektor. Makalah ini akan menganalisis efek kecelakaan water ingress terhadap reaktivitas Doppler teras RGTT200K. Kapabilitas koefisien reaktivitas Doppler untuk mengkompensasi reaktivitas positif yang timbul selama kecelakaan water ingress akan diuji melalui serangkaian perhitungan dengan program MCNPX dan pustaka ENDF/B-VII untuk perubahan temperatur bahan bakar dari 800K hingga 1800K. Tiga opsi kernel bahan bakar UO2, ThO2/UO2 dan PuO2 dengan tiga model kisi bahan bakar pebble di teras reaktor diterapkan untuk kondisi water ingress dengan densitas air dari 0 hingga 1.000 kg/m3. Hasil perhitungan memperlihatkan koefisien reaktivitas Doppler tetap negatif untuk seluruh opsi bahan bakar yang dipertimbangkan bahkan untuk posibilitas water ingress yang besar. Efek water ingress lebih kuat pada model kisi dengan fraksi packing lebih rendah karena lebih banyak volume yang tersedia untuk air yang memasuki teras reaktor. Efek water ingress juga lebih kuat di teras uranium dibandingkan teras thorium dan plutonium sebagai konsekuensi dari fenomena Doppler dimana absorpsi neutron di daerah resonansi 238U lebih besar daripada 232Th dan 240Pu. Secara keseluruhan dapat disimpulkan bahwa, koefisien Doppler teras RGTT200K mampu mengkompensasi insersi reaktivitas yang diintroduksi oleh kecelakaan water ingress. Teras RGTT200K dengan bahan bakar UO2, ThO2/UO2 dan PuO2 dapat mempertahankan fitur keselamatan

  9. Thermal Shock Tests on UO{sub 2} Small Spheres; Essais de choc thermique sur des elements spheriques de UO{sub 2}; Ispytaniya nebol'shikh sharikov iz UO{sub 2} teplovykh udarom; Ensayo de pequenas esferas de UO{sub 2} por choque.termico

    Energy Technology Data Exchange (ETDEWEB)

    Perona, G.; Brutto, E.; Galbusera, U.; Palladino, G.; Sesini, R. [Centro Informazioni Studi Esperienze, Milan (Italy)

    1963-11-15

    exponen los resultados obtenidos. La aplicacion de este metodo presenta, al parecer, considerable interes, sobre todo en lo que concierne a las investigaciones encaminadas a mejorar las caracteristicas de las esferas de UO{sub 2} por medio de aditivos. En e fecto, permite verificar el efecto global con una sola medicion. (author) [Russian] Ispol'zuya malye pariki iz UO{sub 2} v kachestve yadernogo topliva v reaktore, gde oni nakhodyatsya v soprikosnovenii s teplonositelem, neobkhodimo znat' maksimal'nuyu pri rabochem rezhime reaktora velichinu termicheskikh napryazhenij, kotorye mogut vyderzhivat' bez povrezhdeniya ehti shariki. Esli izvestny fizicheskie svojstva materiala, to mozhno rasschitat' ehti napryazheniya pri rabochem rezhime. Odnako vvidu mnogochislennosti podlezhashchikh uchetu faktorov i neizbezhnoj doli neopredelennosti kazhdogo iz nikh predstavlyaetsya predpochtitel'nym provesti neposredstvennye ispytaniya ehtikh sharikov, podvergnuv ikh tem zhe napryazheniyam, kakie oni ispytyvayut v reaktore. V nastoyashchej rabote byl izuchen metod teplovogo udara v primenenii k malym sharikam i ukazyvayutsya usloviya, pri kotorykh ehtot metod pozvolyaet proizvesti napryazheniya, neposredstvenno sravnimye s temi, kotorye sushchestvuyut v reaktore. V sluchae malykh sharikov zatrudnenie zaklyuchaetsya v osushchestvlenii okhlazhdeniya, pozvolyayushchego dostigat' ochen' bol'shikh znachenij koehffitsienta poverkhnostnoj teploperedachi. Opisyvayutsya ehksperimental'nye metody i soobshchayutsya poluchennye rezul'taty. Primenenie ehtogo' metoda, po-vidimomu., predstavlyaet bol'shoj interes, v osobennosti v oblasti tekhnologicheskikh izyskanij s tsel'yu uluchsheniya svojstv malykh sharikov iz UO{sub 2} putem vklyucheniya dobavochnykh komponentov. Fakticheski,ehtot metod daet vozmozhnost' pri pomoshchi odnogo tol'ko izmereniya kontrolirovat' izuchaemoe vozdejstvie. (author)

  10. Slow Neutron Spectrometers at the Swedish Reactors; Spectrometres a Neutrons Lents des Reacteurs Suedois; 0421 041f 0415 041a 0422 0420 041e 041c 0415 0422 0420 042b 041c 0415 0414 041b 0415 041d 041d 042b 0425 041d 0415 0419 0422 0420 041e 041d 041e 0412 041d 0410 0428 0412 0415 0414 0421 041a 0418 0425 0420 0415 0410 041a 0422 041e 0420 0410 0425 ; Espectrometros para Neutrones Lentos en los Reactores de Suecia

    Energy Technology Data Exchange (ETDEWEB)

    Dahlborg, U.; Skoeld, K. [AB Atomenergi, Stockholm (Sweden); Larsson, K. -E. [Royal Institute of Technology, Stockholm (Sweden)

    1965-06-15

    efecto combinado de un filtro de Be y un selector con una curva de transmision estrecha. En este espectrometro, lo mismo que en el de Estocolmo, el selector se coloca delante de la muestra para poder registrar simultaneamente datos a diferentes angulos de observacion. En el reactor R2 se encuentra tambien instalado un espectrometro triaxial de cristal. Los autores describen diversas caracteristicas de los instrumentos, tales como las intensidades y los poderes de resolucion alcanzados, e indican en que medida se adaptan a ciertas determinaciones. Demuestran con datos numericos que un aumento relativamente pequeno del poder de resolucion ocasiona una elevada perdida de intensidad. Cuando se comparan los reactores RI y R2 como fuentes neutronicas para experimentos en el orificio de haz, es interesante observar que, si bien el flujo neutronico del reactor R2 es 100 veces mayor, el rendimiento en neutrones es unas 10 veces menor. El fenomeno se debe a la estrechez de los orificios de haz y a los filtros que son necesarios para reducir los flujos gamma y de neutrones rapidos. Como ilustracion, la memoria discute brevemente datos de dispersion en H{sub 2}O obtenidos con los diversos instrumentos. La comparacion demuestra que, para experimentos con el orificio de haz, el reactor de agua pesada presenta ventajas innegables. (author) [Russian] V nastojashhee vremja na dvuh shvedskih issledovatel'skih reaktorah K1 v Stokgol'me i R2 v Studsvike, imejutsja vozmozhnosti dlja ispol'zovanija chetyreh razlichnyh spektrometrov nejtronov v opytah po neuprugomu rassejaniju nejtronov. V Stokgol'me na tjazhelovodnom reaktore R1 moshhnost'ju 600-kvt odnovremenno ispol'zujutsja dva spektrometra po vremeni proleta s medlenno dejstvujushhim preryvatelem. Na odnom reaktore my postojanno ispol'zuem berillievyj fil'tr v kachestve monohromatora, v to vremja kak na drugom reaktore mozhno ispol'zovat' libo berillievyj fil'tr, libo kristallicheskij monohromator. Obnaruzheno, chto pri izmerenijah

  11. A Survey of the Fuel Cycles Operated in the United Kingdom; Etude d'ensemble sur les cycles de combustible au Royaume-Uni; Obzor toplivnykh tsiklov, ispol'zuemykh v soedinennom korolevstve; Estudio de los ciclos de combustible utilizados en el Reino Unido

    Energy Technology Data Exchange (ETDEWEB)

    Allday, C. [United Kingdom Atomic Energy Authority, Risley, Warrington, Lancs (United Kingdom)

    1963-10-15

    enriquecido tambien se puede utilizar como combustible oxido de uranio natural enriquecido con plutonio. En la memoria se resume la experiencia adquirida en la produccion de combustible de oxido para el AGR y en la explotacion del reactor y los planes para la regeneracion del combustible. Se examina la posibilidad de utilizar combustible de plutonio y se analizan las consecuencias que tendria su adopcion sobre los costos y el ciclo del combustible. Por ultimo, se destaca la importancia de los reactores Magnox y AGR en el programa energetico del Reino Unido. (author) [Russian] a ) Prirodnyj uran/topdivnyj tsikl ''Magnoks''. Soedinennoe Kor olevstvo izb ralo reaktor na prirodnom urane s grafitovym zam edli tel em i gazovy m okhlazhdeniem v kachestve osnovy programmy po yadernoj ehnergii. Ono ehkspluatirovalo ehti reaktory v Kolder-Kholle i Chepelkrosse v techenie semi det; reaktory v Berkli i Braduehlle v nastoyashchee vremya nakhodyatsya v stadii ehkspluatatsii, a reaktory v semi drugikh mestakh v stadii stroitel'stva ili planirovaniya. Toplivo dlya ehtikh reaktorov proizvoditsya na zavode v Springfilde i zatem perevozitsya dlya zagruzki k mestopolozheniyu reaktora. Posle oblucheniya i razgruzki toplivo transportiruetsya na zavod v Uindskejl dlya otdeleniya urana i plutoniya ot produktov deleniya. Daetsya opisanie opyta CK v oblasti konstruktsii i proizvodstva toplivnykh ehlementov, ehkspluatatsii reaktora, transportirovki obluchennogo topliva i posleduyushchej obrabotki topliva. Upominaetsya o povedenii topliva v reaktore i ob al'ternativnykh programmakh zagruzki l razgruzki toplivnykh ehlementov; ehta tema razrabatyvaetsya v drugikh trudakh. b) Reaktory, ispol'zuyushchie obogashchennoe toplivo. Soedinennoe Korolevstvo razrabatyvaet usovershenstvovannyj reaktors gazovym okhlazhdeniem AGE, prototip kotorogo voshel v stroj v 1963 godu. Toplivo proizvoditsya iz obogashchennoj okisi urana, zaklyuchennoj v obolochku iz nerzhaveyushchej stali, i Sudet pererabatyvat'sya posredstvom

  12. Preparation of Impervious Pyrolytic Carbon Coatings and Application to Dispersed Fuels; Preparation de revetements de carbone pyrolytique etanches - applications aux combustibles disperses; Prigotovlenie nepronitsaemogo uglerodnogo piroliticheskogo pokrytiya dlya dispergirovannogo topliva; Preparacion de revestimientos estancos de carbono piroutico: aplicacion a los combustibles nucleares dispersos

    Energy Technology Data Exchange (ETDEWEB)

    Auriol, A.; David, C. [Battelle Memorial Institute, Geneve (Switzerland); Fillatre, A.; Kurka, G.; Le Boulbin, E.; Rappeneau, J. [Commissariat a l' Energie Atomique (France)

    1963-11-15

    osadka. Ehtot sposob pokrytiya byl perenesen na zerna okisi i karbida urana metodom dvizhushchegosya sloya. Posle utochneniya uslovij pokrytiya ehtikh zeren byla issledovana ikh makro- i mikrostruktury, a takte ikh pronitsaemost'. Byli izucheny svojstva ehtikh zeren pri vysokoj temperature na predmet ikh vozmozhnogo primeneniya v reaktore. (author)

  13. Measurement of nuclear reactor noise at low power levels; Merenje nuklearnog reaktorskog suma na malim snagama

    Energy Technology Data Exchange (ETDEWEB)

    Velickovic, Lj; Petrovic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1968-11-15

    mogu da se interpretiraju pomocu prostorno-nezavisne reaktorske teorije. Fizicki analiza reaktorskog suma se ogranicila na odredjivanje odnosa {beta}/l iz autokorelaeionih merenja na snazi od 0,5 W (reaktor 'RB' u Vinci). Analizirane su tri razlicite konfiguracije goriva (resetke koraka 8 cm, 11,3 cm i 14 cm). Pokazano je da je moguce odrediti parametar {beta}/l iz autokorelacionih merenja neutronske gustine sa velikom tacnoscu (nekoliko %). (author)

  14. Produksi Biofuel dari Minyak Kelapa Sawit dengan Katalis Au/HZSM-5 dan Kompositnya

    Directory of Open Access Journals (Sweden)

    Tillotama Anindita Sari

    2012-09-01

    diperoleh untuk katalis Au/HZSM-5 yield tertinggi kerosene 25,24%, gasoline 15,69% dan diesel 10,71% pada temperatur reaktor 500 °C dengan laju alir  gas N2 90 ml/min. Untuk katalis Komposit (HZSM-5/MCM-41 yield tertinggi diesel 26,53%, kerosene 19,26% dan gasoline 6,41% pada temperatur 450 °C laju alir 300 ml/min serta pada temperatur 350 °C dengan laju alir 90 ml/min dengan yield diesel tertinggi 24,38%, kerosene 18,84% dan gasoline 4,41%.

  15. PSA LEVEL 3 DAN IMPLEMENTASINYA PADA KAJIAN KESELAMATAN PWR

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    Full Text Available Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahwa implementasi metode kajian keselamatan PSA Level 3 sangat efektif dan komprehensif terhadap estimasi dampak, konsekuensi, risiko, kesiapsiagaan kedaruratan nuklir (nuclear emergency preparedness, dan manajemen kecelakaan reaktor terutama untuk kecelakaan parah atau kecelakaan luar dasar desain PLTN. Hasil kajian dapat digunakan sebagai umpan balik untuk kajian keselamatan PSA Level 1 dan PSA Level 2. Kata kunci: PSA level 3, kecelakaan, PWR   Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidentsof PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The

  16. Hydraulic Behavior in The Downflow Hanging Sponge Bioreactor

    Directory of Open Access Journals (Sweden)

    Izarul Machdar

    2016-12-01

    Full Text Available Performance efficiency in a Downflow Hanging Sponge (DHS bioreactor is associated with the amount of time that a wastewater remains in the bioreactor. The bioreactor is considered as a plug flow reactor and its hydraulic residence time (HRT depends on the void volume of packing material and the flow rate. In this study, hydraulic behavior of DHS bioreactor was investigated by using tracer method. Two types of sponge module covers, cylindrical plastic frame (module-1 and plastic hair roller (module-2, were investigated and compared. A concentrated NaCl solution used as an inert tracer and input as a pulse at the inlet of DHS bioreactor. Analysis of the residence time distribution (RTD curves provided interpretation of the index distribution or holdup water (active volume, the degree of short-circuiting, number of tanks in series (the plug flow characteristic, and the dispersion number. It was found that the actual HRT was primarily shorter than theoretical HRT of each test. Holdup water of the DHS bioreactor ranged from 60% to 97% and 36% to 60% of module-1 and module-2, respectively. Eventhough module-1 has higher effective volume than module-2, result showed that the dispersion numbers of the two modules were not significant difference. Furthermore, N-values were found larger at a higher flow rate. It was concluded that a DHS bioreactor design should incorporated a combination of water distributor system, higher loading rate at startup process to generate a hydraulic behavior closer to an ideal plug flow.ABSTRAKEfisiensi unjuk kerja bioreactor Downflow Hanging Sponge (DHS berkaitan dengan lamanya waktu tinggal limbah berada di dalam bioreaktor tersebut. Bioreaktor DHS dianggap sebagai seuatu reaktor aliran sumbat (plug flow dimana waktu tinggal hidraulik (HRT tergantung pada volume pori material isian dan laju alir. Dua jenis modul digunakan dalam penelitian ini, yang diberi nama dengan module-1 dan module-2 untuk melihat pengaruh jenis modul

  17. Demolition of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklearservice, Essen (Germany); Cremer, J. [SNT Siempelkamp Nukleartechnik, Heidelberg (Germany)

    2003-06-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [German] Der mit Leichtwasser gekuehlte und moderierte Schwimmbad-Forschungsreaktor FRJ-2 (MERLIN) wurde von 1958 bis 1962 fuer die damalige Kernforschungsanlage Juelich (KFA) errichtet. Von 1964 bis 1985 wurde er fuer Experimente mit zunaechst 5 MW und spaeter 10 MW thermischer Leistung bei einem maximalen thermischen Neutronenfluss von 1,1.10{sup 14} n/cm{sup 2}s genutzt. Im Jahr 1985 stellte der Reaktor seinen Betrieb ein. Die Brennelemente wurden aus der Anlage entfernt und in die USA und nach Grossbritannien verbracht. Seit 1996 erfolgen die wesentlichen Abbautaetigkeiten unter Leitung eines verantwortlichen Projektteams. Bis Ende 1998 wurde das komplette Sekundaerkuehlsystem entfernt. Dem Abbau der Kuehlkreislaeufe und Experimentiereinrichtungen folgte im Jahr 2000 der Ausbau der

  18. ANALISIS TRANSIEN PADA PASSIVE COMPACT MOLTEN SALT REACTOR (PCMSR

    Directory of Open Access Journals (Sweden)

    M. Makrus Imron

    2015-04-01

    Full Text Available Penggunaan bahan bakar cair berupa garam LiF-BeF2-ThF4-UF4 pada Passive Compact Molten Salt Reactor (PCMSR meyebabkan pengendalian daya pada PCMSR dapat dilakukan dengan mengendalikan laju aliran bahan bakar dan pendingin. Sedangkan dari sistem keselamatan, penggunaan bahan bakar cair menjadikan PCMSR memiliki karakter keselamatan melekat (inherent safety yang baik. Pada penelitian ini telah dilakukan analisis transien PCMSR pada tiga kondisi, yaitu: ketika terjadi perubahan laju aliran bahan bakar, ketika terjadi perubahan laju aliran pendingin dan ketika terdapat kegagalan pada sistem pelepasan panas (loss of heat sink. Penelitian dilakukan dengan memodelkan reaktor pada kondisi tunak menggunakan paket program. Standart Reactor Analysis Code (SRAC. Selanjutnya dari keluaran paket program SRAC diperoleh data data yang meliputi fluks netron,konstanta grup, kontanta peluran prekusor netron, fraksi netron kasip untuk perhitungan transien. Penelitian ini menunjukkan bahwa penurunan laju aliran bahan bakar sebesar 50 % dari laju bahan bakar sebelumnya, menyebabkan daya pada PCMSR turun menjadi 78 % dari daya sebelumnya. Dan penurunan laju aliran pendingin sebesar 50 % dari laju pendingin sebelumnya, menyebabkan daya pada PCMSR turun menjadi 63 % dari daya sebelumnya. Sedangkan pada saat terjadi loss of heat sink daya PCMSR menunjukkan penurunan. Kata kunci: PCMSR, transien, daya, laju aliran.   The use of liquid fuels in the form of molten salts LiF-BeF2-ThF4-UF4 in Passive Compact Molten Salt Reactor (PCMSR makes power control at PCMSR can be done by controlling the flow rate of fuel and coolant. In addition, from safety systems aspect, the use of liquid fuels makes PCMSR has good inherent safety characteristics. In this study transient analysis has been carried out on three conditions of PCMSR, namely when the fuel flow rate is changing, when the coolant flow rate is changing and when there is loss of heat sink condition. This research is

  19. Performance Characteristics of the Experimental Boiling Water Reactor from 0 to 100 MW(t); Performances de l'EBWR de 0 a 100 MW; Rabochaya kharakteristika ehksperimental'nogo kipyashchego reaktora EBWR pri moshchnosti 0 - 100 mgvt.; Rendimiento del reactor experimental de agua hirviente (EBWR) entre 0 y 100 MW

    Energy Technology Data Exchange (ETDEWEB)

    Iskenderian, A.; Lipinski, W. C.; Petrick, M.; Wimunc, E. A. [Argonne National Laboratory, Argonne, IL (United States)

    1963-10-15

    entonces de comportarse como reactor de agua hirviente de ciclo directo; en cierto modo, funciona como reactor de ciclo doble y circulacion natural. (author) [Russian] 25 maya 1962 goda Argonnskaya natsional'naya laboratoriya poluchila razreshenie KAEH SSHA na ehkspluatatsiyu reaktora EBWR na moshchnosti 100 mgvt. Administrativnoe razreshenie na ehkspluatatsiyu reaktora bylo predostavleno sistemoj garantij. Mezhdunarodnogo agentstva po atomnoj ehnergii 11 iyulya 1961 goda. 15 noyabrya 1962 goda byl dostignut uroven' moshchnosti v 100 mgvt. 6 dekabrya 1962 goda ehksperimental'naya programma byla zakonchena. Odnoj iz osnovnykh tselej ee byla tshchatel'naya proverka reaktora dlya polucheniya dannykh i informatsii rabochej kharakteristiki ehtogo tipa reaktora. Ehta programma byla pervoj programmoj takogo roda i pervoj vypolnennoj programmoj. Dlya polucheniya nuzhnykh dannykh neobkhodimo bylo razrabotat' mnogie novye pribory. TSel' byla dostignuta, polucheno mnogo novykh dannykh o rabochej kharakteristike kipyashchego reaktora s estestvennoj tsirkulyatsiej. Tak,naprimer, poluchena informatsiya otnositel'no skorosti potoka tsirkulyatsii v zamknutom tsikle, predelov separatsii zhidkogo para (vydelenie para v osadok v spusknoj trube i unos zhidkosti ehflu- entom para), nedogreva, lokalizatsii dejstvitel'noj poverkhnosti razdela v reaktore i ee svyazi s urovnem vodnoj kolonki, skorosti razrusheniya para v spusknoj trube, pustotnykh koehffitsientov, reaktivnoj sposobnosti H{sub 3}BO{sub 3}, temperaturnykh koehffitsientov, ispol'zovaniya sterzhnej iz bora dlya tselej kontrolya, ispol'zovaniya svezhikh toplivnykh ehlementov, peredatochnykh funktsij,analiza shuma, nekotorykh izmerenij potoka, stabil'nosti i t.d. Krome togo, byli polucheny dannye o povedenii i tselostnosti nekotorykh reaktornykh komponentov i sistem, takikh, kak bornokislaya kontrol'naya reaktsiya, urovni radiatsii, raspredelenie produktov korrozii, vykhod iz stroya oborudovaniya, toplivo i reguliruyushchie sterzhni i t

  20. APPLICATION OF PHYTOREMEDIATION FOR HERBAL MEDICINE WASTE AND ITS UTILIZATION FOR PROTEIN PRODUCTION

    Directory of Open Access Journals (Sweden)

    Danny Soetrisnanto

    2012-11-01

    bertujuan untuk mengevaluasi penggunaan tanaman air (enceng gondok dan teratai untukmereduksi kontaminan dalam limbah obat jamu. Phytoremediasi dilakukan selama 4-8 hari dan tinggicairan dalam reaktor dijaga pada 5 cm. Keluaran dari phytoremediasi pertama menggunakantanaman air digunakan sebagai medium di phytoremediasi menggunakan mikroalga Spirulina. Untukmendapakan pertumbuhan yang optimum, maka ditambahakan juga nutrient dan menunjukkan bahwaSpirulina tumbuh dengan sangat baik dalam medium ini. Pertumbuhan terbaik diperoleh dariphytoremediasi menggunakan teratai selama 3 hari dan kecepatan pertumbuhan 0,271/hari denganperbandingan C:N:P = 57,790:9,28:1.

  1. Nonlinear dynamics in chemical processes. Project A: Locally distributed periodic processes. Sub-project A3I: Catalitic afterburning. Nonlinear periodic front travelling processes. Final report; Nichtlineare Dynamik bei chemischen Prozessen. Projekt A: Oertlich verteilte periodische Prozesse. Teilprojekt A3I: Katalytische Nachverbrennung im Zirkulationsreaktor. Nichtlineare periodische Frontwanderungsprozesse. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Richter, M.; Reinhardt, H.J.; Roschka, E.

    1998-01-31

    Massnahmen zur Prozessfuehrung kann das ortsabhaengige Temperaturmaximum der zirkulierenden Reaktionsfront innerhalb der katalysatorspezifischen Grenztemperaturen fuer Totaloxidation bzw. Katalysatorschaedigung gehalten werden. Aus den Untersuchungen resultiert ein anwendungsbereites Konzept fuer den Einsatz des Zirkulationsreaktors zur Zersetzung schwerspaltbarer Schadstoffe. Vorauszusetzen ist die Verfuegbarkeit eines geeigneten Katalysators. Als bevorzugter Einsatzfall wird die katalytische Nachverbrennung wechselnder Schadstoffmengen in kleinen Abgasstroemen betrachtet, um das autonome Einschwingen des Reaktors in einen neuen Betriebszustand bei veraenderten Eingangsbedingungen vorteilhaft zu nutzen. (orig.)

  2. Breeder development and breeder strategies worldwide; Brueterentwicklung und Brueterstrategien weltweit

    Energy Technology Data Exchange (ETDEWEB)

    Marth, W. [Stabsabteilung Finanzen/Controlling, Bereich Stillegung Nuklearer Anlagen, Forschungszentrum Karlsruhe GmbH (Germany)

    1997-05-01

    indische Versuchsbrueter FBTR ist in der Inbetriebnahmephase von mancherlei technischen Problemen geplagt. Der amerikanische Schnelle Reaktor FFT wird seit Jahren im heissen, aber nichtnuklearen Zustand gehalten, da sich die Politiker ueber seine Bestimmung nicht einigen koennen. Das Schicksal des deutschen Brueterprototyps SNR-300 ist bekannt: Dem Kernkraftwerk Kalkar wurde sechs Jahre lang die Betriebsgenehmigung versagt. Nach seiner politisch erzwungenen Stillegung im Jahre 1991 wurde es teilweise demontiert und in einen Freizeitpark umgewandelt. (orig.)

  3. Annual Report 2007 VUJE

    International Nuclear Information System (INIS)

    Anon

    2008-01-01

    VUJE foundation is closely connected with the history of Bohunice nuclear power plant. The team of research and development employees in Bohunice nuclear power plant formed the basis for foundation of the Vyskumny ustav jadrovych elektrarni (Nuclear Power Plant Research Institute - hereinafter referred to as VUJE or Company). Institute started its operations on January 1, 1977 as a concern company of Slovenske energeticke podniky (Slovak Power Enterprises). One year later VUJE became an independent organization controlled directly by the Federal Ministry of Fuels and Power. Excellent results were the impulse for the federal government to accredit VUJE in 1983 to lead the scientific commissioning of all nuclear power plants in Czechoslovakia. In 1983 started VUJE the branch education and training centre, which was in the next year equipped with a full-range simulator of WWER 440/213 reaktor unit. This centre gradually achieved international recognition. The international Atomic Energy Agency in Vienna also carried out training courses here. VUJE moved its headquarters to Trnava in 1985, however multiple experimental facilities, development workshops and some laboratories stayed in the area of nuclear power plants in Jaslovske Bohunice. At the end of the eighties VUJE counted 759 employees and gained an important position not only in Czechoslovakia, but also within international nuclear power industry. Important changes in the whole society in 1989 also influenced the future in the institute. Management developed a new organizational model of VUJE and step-by- step also the institute's strategy and line of business. Company retained its focus and was transformed in to a state company. In the following privatisation the joint-stock company established by the institute's employees bought the state research institute. The institute was transformed into a private joint-stock company Vyskumny ustav jadrovych elektrarni Trnava, a.s as of November f 1, 1994. A The

  4. Annual report 2006 VUJE

    International Nuclear Information System (INIS)

    Anon

    2006-01-01

    VUJE foundation is closely connected with the history of Bohunice nuclear power plant. The team of research and development employees in Bohunice nuclear power plant formed the basis for foundation of the Vyskumny ustav jadrovych elektrarni (Nuclear Power Plant Research Institute - hereinafter referred to as VUJE or Company). Institute started its operations on January 1, 1977 as a concern company of Slovenske energeticke podniky (Slovak Power Enterprises). One year later VUJE became an independent organization controlled directly by the Federal Ministry of Fuels and Power. Excellent results were the impulse for the federal government to accredit VUJE in 1983 to lead the scientific commissioning of all nuclear power plants in Czechoslovakia. In 1983 started VUJE the branch education and training centre, which was in the next year equipped with a full-range simulator of WWER 440/213 reaktor unit. This centre gradually achieved international recognition. The international Atomic Energy Agency in Vienna also carried out training courses here. VUJE moved its headquarters to Trnava in 1985, however multiple experimental facilities, development workshops and some laboratories stayed in the area of nuclear power plants in Jaslovske Bohunice. At the end of the eighties VUJE counted 759 employees and gained an important position not only in Czechoslovakia, but also within international nuclear power industry. Important changes in the whole society in 1989 also influenced the future in the institute. Management developed a new organizational model of VUJE and step-by- step also the institute's strategy and line of business. Company retained its focus and was transformed in to a state company. In the following privatisation the joint-stock company established by the institute's employees bought the state research institute. The institute was transformed into a private joint-stock company Vyskumny ustav jadrovych elektrarni Trnava, a.s as of November f 1, 1994. A The

  5. Forecasting the Quantity and Activity of Fission Products in France in Future Years in the Light of Atomic Energy Development; Quantite et Activite des Produits de Fission Obtenus en France dans les Annees a Venir Compte Tenu du Developpement de l'Energie Atomique; 041a 041e 041b 0414 ; Cantidad y Actividad de los Productos de Fision que se Obtendran en Francia en los Anos Venideros, Habida Cuenta del Desarrollo de la Energia Atomica

    Energy Technology Data Exchange (ETDEWEB)

    Guirlet, J.; Lavie, J. M. [Commissariat a l' Energie Atomique, Saclay (France)

    1960-07-01

    en un momento dado, asi como la actividad de los productos de fision para un periodo determinado. (author) [Russian] Pri pomoshhi formuly Vignera-Uoja teoreticheski vozmozhno predusmotret' aktivnost' slozhnoj smesi produktov raspada ljubogo reaktora. Issledovanija provodilis' s uchetom vozmozhnogo razvitija atomnoj jenergetiki vo Francii do 1975 goda. Predpolagalos', chto uran v reaktore budet nahodit'sja tri ili shest' mesjacev. Takzhe vozmozhno ustanovit' aktivnost' opredelennogo produkta raspada i opredelit' ego radioaktivnyj raspad. Pri issledovanijah dlja trehmesjachnoj aktivacii byl vybran stroncij. Kazhdyj kompleks grafikov daet obshhuju aktivnost' dlja ljubogo perioda, a takzhe, sobstvennuju aktivnost' produktov raspada, sootvetstvujushhuju opredelennomu periodu. (author)

  6. Karakterisasi Paduan AlMgSi Untuk Kelongsong Bahan Bakar U3Si2/Al Dengan Densitas Uranium 5,2 gU/cm3

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    Aslina Br. Ginting

    2018-03-01

    Full Text Available Meningkatnya densitas uranium dari 2,96 gU/cm3 menjadi 5,2 gU/cm3 bahan bakar U3Si2/Al harus diikuti dengan penggunaan kelongsong yang kompatibel. Bahan bakar berdensitas tinggi mempunyai kekerasan yang tinggi, sehingga bila menggunakan paduan AlMg2 sebagai kelongsong dapat menyebabkan terjadi dogbone pada saat perolan. Selain fenomena dogbone, pada saat bahan bakar tersebut digunakan di reaktor dapat terjadi swelling karena meningkatnya hasil fisi maupun burn up. Oleh karena itu, perlu dicari pengganti bahan kelongsong untuk bahan bakar U3Si2/Al densitas tinggi. Pada penelitian ini telah dilakukan karakterisasi paduan AlMgSi sebagai kandidat pengganti kelongsong AlMg2. Karakterisasi yang dilakukan meliputi analisis termal, kekerasan, mikrostruktur dan laju korosi. Analisis termal dilakukan menggunakan DTA (Differential Thermal Analysis dan DSC (Differential Scanning Calorimetry. Analisis kekerasan menggunakan alat uji kekerasan mikro, mikrostruktur menggunakan SEM (Scanning Electron Microscope dan analisis laju korosi dilakukan dengan pemanasan pada temperatur 150 oC selama 77 jam di dalam autoclave. Hasil analisis menunjukkan bahwa kelongsong AlMgSi maupun AlMg2 mempunyai kompatibilitas panas dengan bahan bakar U3Si2/Al cukup stabil hingga temperatur 650 oC. Kelongsong AlMgSi mempunyai kekerasan sebesar 115 HVN dan kelongsong AlMg2 sebesar 70,1 HVN. Sementara itu, analisis mikrostruktur menunjukkan bahwa morfologi ikatan antarmuka (interface bonding kelongsong AlMgSi lebih baik dari kelongsong AlMg2, demikian halnya dengan laju korosi bahwa kelongsong AlMgSi mempunyai laju korosi lebih kecil dibanding kelongsong AlMg2. Hasil karakterisasi termal, kekerasan, mikrostruktur dan laju korosi menunjukkan bahwa PEB U3Si2/Al densitas 5,2 gU/cm3 menggunakan kelongsong AlMgSi lebih baik dibanding PEB U3Si2/Al  densitas 5,2 gU/cm3  menggunakan kelongsong AlMg2. Kata kunci: U3Si2/Al, densitas 5,2 gU/cm3, kelongsong AlMgSi dan AlMg2.

  7. Tingkat Keamanan Konsumsi Residu Karbamat dalam Buah dan Sayur Menurut Analisis Pascakolom Kromatografi Cair Kinerja Tinggi

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    Bambang Wispriyono

    2013-02-01

    Full Text Available Karbamat merupakan salah satu jenis pestisida yang banyak digunakan untuk membasmi hama buah dan sayur. Untuk menentukan bahwa residu karbamat dalam sayuran masih aman dikonsumsi manusia, telah dilakukan analisis beberapa residu karbamat seperti metomil, karbaril, karbofuran, dan propoksur. Sampel-sampel tomat, apel, selada air, kubis, dan sawi hijau dikumpulkan dari tiga supermarket dan satu pasar tradisional di Depok, Jawa Barat. Analisis dilakukan serempak untuk ke empat residu karbamat menggunakan kromatografi cair kinerja tinggi denganpereaksi o-ftalaldehida dan 2-merkaptoetanol dalam reaktor pascakolom dengan detektor fluoresensi. Dari sampel-sampel buah dan sayur yang dianalisis, hanya sawi hijau asal pasar tradisional yang positif mengandung propoksur dengan kadar 1,2 mg/25 gram berat basah (0,048 mg/g berat basah. Dengan Acceptable Daily Intake(ADI propoksur 0,005 mg/kg berat badan/hari, konsumsi sawi hijau harian seberat 20 g/hari masih cukup aman dari gangguan kesehatan akibat pajanan kronik propoksur dengan margin of safety 298,7 (> 100 sebagai batas aman. Carbamat is a group of pesticides which is commonly used to control fruits and vegetables pests. To determine that carbamat residues in fruits and vegetables are safe for human consumption, carbamate residues such as methomyl, carbaryl, carbofuran, and propoxur in vegetables and fruits have been analyzed. Samples of tomato, apple, water lettuces, cabbage, and mustard greens were collected from three supermarkets and one traditional market in Depok, West Java. The analysis was carried out simultaneously for all four carbamate residues by high performance liquid chromatography using o-phtaladehyde and 2 mercaptoethanol reagents in post-column reactor with a fluorescence detector. Of fruits and vegetable samples analyzed, only mustard greens from traditional market positively containe propoxur at 1.2 mg/ 25 gram wet weight (0,048 mg/gram wet weight. With Acceptable Daily Intake (ADI

  8. STUDI EKSPERIMEN PEMILIHAN BIOMASSA UNTUK MEMPRODUKSI GAS ASAP CAIR ( LIQUID SMOKE GASES SEBAGAI BAHAN PENGAWET

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    Sugeng Slamet

    2015-04-01

    Full Text Available ABSTRAK Pengertian umum asap cair merupakan suatu hasil destilasi atau pengembunan dari uap hasil pembakaran tidak langsung maupun langsung dari bahan yang banyak mengandung karbon dan senyawa- senyawa lain. Bahan baku yang banyak digunakan untuk membuat asap cair adalah kayu, bongkol kelapa sawit, ampas hasil penggergajian kayu, dan lain-lain. Pembuatan asap cair menggunakan metode pirolisis yaitu peruraian dengan bantuan panas tanpa adanya oksigen atau dengan jumlah oksigen yang terbatas. Biasanya terdapat tiga produk dalam proses pirolisis yakni: gas, pyrolisis oil, dan arang, yang mana proporsinya tergantung dari metode pirolisis, karakteristik biomassa dan parameter reaksi. Metode yang dilakukan diawali dengan melakukan rancang bangun unit pirolisator lengkap dengan perangkat kondensor dengan pipa tembaga tipe spiral untuk memproduksi gas asap cair dari bahan biomassa yang dipilih yaitu tempurung kelapa dan sampah organik. Metode Pirolisis yang merupakan proses reaksi penguraian senyawa-senyawa penyusun kayu keras menjadi beberapa senyawa organik melalui reaksi pembakaran kering pembakaran tanpa oksigen. Reaksi ini berlangsung pada reaktor pirolisator dengan variasi temperatur 150oC, 250oC dan 300oC selama 8 jam pembakaran. Asap hasil pembakaran dikondensasi dengan kondensor yang berupa pipa tembaga melingkar. Hasil dari proses pirolisis diperoleh tiga produk yaitu asap cair, tar, dan arang. Kondensasi dilakukan dengan pipa atau koil melingkar yang dipasang dalam bak pendingin. Air pendingin dapat berasal dari air hujan yang ditampung dalam bak penampungan. Hasil yang diperoleh dari penelitian ini adalah biomassa tempurung kelapa menghasilkan jumlah senyawa fenol lebih besar 30-33%. Hal ini menunjukkan bahwa pada jenis biomassa ini lebih unggul dalam fungsi sebagai antioksidan, karena kaya akan kandungan senyawa fenol, sehingga lebih optimal dalam hal menghambat kerusakan pangan dengan cara mendonorkan hidrogen. Sedangkan biomassa cangkang

  9. Surfaktan Sodium Ligno Sulfonat (SLS dari Debu Sabut Kelapa

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    Mukti Mulyawan

    2015-03-01

    Full Text Available Indonesia merupakan negara agraris yang menghasilkan beragam hasil pertanian yang melimpah. Salah satu hasil pertanian yang menonjol di Indonesia adalah kelapa. Produksi buah kelapa di Indonesia rata-rata sebanyak 15,5 miliar butir/tahun atau setara dengan 3,02 juta ton kopra, 3,75 juta ton air, 0,75 juta ton arang tempurung, 1,8 juta ton serat sabut (coir fiber dan 3,3 juta ton debu sabut (coir dust/ cocopeat. Komposisi sabut kelapa terdiri dari 25% gabus dan 75% serat . Tetapi, debu sabut kelapa masih dikembangkan sebatas sebagai media tanam. sisanya akan menjadi limbah dengan kontribusi sangat besar dari pengisi pada volume total sampah domestiK. Banyaknya komoditas kelapa dan potensi limbah sabut yang dihasilkan, membuat pemanfaatan Debu Sabut menjadi bahan yang bernilai ekonomis patut untuk dilakukan. Salah satunya adalah sebagai bahan pembuatan Surfakatan Sodium Ligno Sulfonat (SLS yang selama ini komoditasnya diperoleh seluruhnya dari impor. Adapun tahapan proses pembuatan SLS dari Debu Sabut kelapa adalah mempersiapan Bahan Baku berupa Debu Sabut Kelapa. Dilanjutkan dengan pemasakan/pulping menggunakan metode organosolv dengan alat pemasak digester (R-120. Dari lindi hitam yang dihasilkan, akan diproses dengan Isolasi Lignin dengan metode presipitasi asam. Lindi hitam yang telah didapat diendapkan dengan menambahkan secara perlahan H2SO4dengan konsentrasi 20% sampai pH 2 pada tangki isolasi pertama (M-211.Proses isolasi dengan metode pengasaman banyak digunakan untuk mendapatkan lignin dengan kemurnian tinggi. Untuk Menghasilkan SLS, Lignin Isolat perlu direaksikan dengan bahan penyulfonasi natrium bisulfit (NaHSO3, sehingga menghasilkan natrium lignosulfonat (SLS pada reaktor sulfonasi (R-310. Berlokasi di Provinsi Riau, Pabrik ini akan dibangun dengan kapasistas 20.150 ton/tahun. Dari analisa ekonomi, diperlukan Modal tetap (FCI sebesar Rp 316.323.349.677; Modal kerja (WCI sebesar Rp 74.429.023.453; Investasi total (TCI sebesar Rp 390

  10. Pengaruh Kecepatan Homegenisasi Terhadap Sifat Fisika dan Kimia Krim Nanopartikel dengan Metode High Speed Homogenization (HSH

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    Galuh Suprobo

    2015-06-01

    Full Text Available Nanoparticle cream is the development of nanotechnology in cosmetics fields for improving the function of cream. High speed homogenization (HSH is one of the methods for creating nanoparticle cream. In this research, the use of natural materials based palm oil derivative  such as stearic acid, cetil alcohol, cetil stearil alcohol was chosen in nanoparticle cream producing by using HSH methods.The speed variable of  homogenization of 1000 rpm, 1500 rpm, 2,000 rpm and 2,500 rpm  intended to find out the influence of speed toward the  properties of cream product. The observation result showed the influence on physical display in term of texture but not in homogeneity , stability and cream color. The pH of the product during two months storage for all variables were still stable. The particle size was increased in the homogeneity of speed at 2000 rpm and 2500 rpm. In this research has produced the cream in particle size from 239.86 to 358.10 nm which enter in nanoparticle category 50 nm to 1000 nm. The stability of nanoparticle cream product in the range of 97,20 to 98%.ABSTRAKKrim nanopartikel merupakan pengembangan nanoteknologi di bidang kosmetik untuk meningkatkan fungsi krim tersebut. High speed homogenization (HSH merupakan salah satu metoda dalam pembuatan krim nanopartikel. Pada penelitian ini, krim nanopartikel dibuat menggunakan bahan baku alami turunan kelapa sawit yaitu asam stearat, setil alkohol, setil stearil alkohol dengan metoda HSH. Variabel kecepatan homogenisasi pada 1000 rpm, 1500 rpm, 2000 rpm dan 2500 rpm dimaksudkan untuk mengetahui pengaruh kecepatan terhadap sifat-sifat krim. Hasil menunjukkan bahwa perubahan kecepatan homogenisasi dalam reaktor berpengaruh terhadap tampilan fisik dari segi tekstur, akan tetapi tidak mempengaruhi terhadap kehomogenan, stabilitas dan warna krim. Dari pengamatan selama 2 bulan penyimpanan diketahui tidak terjadi perubahan pH selama penyimpanan untuk keempat variabel. Ukuran partikel

  11. TRANSESTERIFICATION OF VEGETABLES OIL USING SUBAND SUPERCRITICAL METHANOL

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    Nyoman Puspa Asri

    2012-11-01

    Full Text Available A benign process, non catalytic transesterification in sub and supercritical methanol method was usedto prepare biodiesel from vegetables oil. The experiment was carried out in batch type reactor (8.8 mlcapacity, stainless steel, AKICO, JAPAN by changing the reaction condition such as reactiontemperature (from 210°C in subcritical condition to 290°C in supercritical state with of 20°Cinterval, molar ratio oil to methanol (1:12-1:42 and time of reaction (10-90 min. The fatty acidmethyl esters (FAMEs content was analyzed by gas chromatography-flame ionization detector (GCFID.Such analysis can be used to determine the biodiesel yield of the transesterification. The resultsshowed that the yield of biodiesel increases gradually with the increasing of reaction time atsubcritical state (210-230oC. However, it was drastically increased at the supercritical state (270-290oC. Similarly, the yield of biodiesel sharply increased with increasing the ratio molar of soy oilmethanolup to 1:24. The maximum yield 86 and 88% were achieved at 290oC, 90 min of reaction timeand molar ratio of oil to methanol 1:24, for soybean oil and palm oil, respectively.Proses transesterifikasi non katalitik dengan metanol sub dan superkritis,merupakan proses yang ramah lingkungan digunakan untuk pembuatan biodiesel dari minyak nabati.Percobaan dilakukan dalam sebuah reaktor batch (kapasitas 8,8 ml, stainless steel, AKICO, JAPAN,dengan variabel kondisi reaksi seperti temperatur reaksi (dari kondisi subkritis 210°C-kondisisuperkritis 290°C dengan interval 20°C, rasio molar minyak-metanol (1:12-1:42 dan waktu reaksi(10-90 menit. Kandungan metil ester asam lemak (FAME dianalisis dengan kromatografi gasdengan detektor FID (GC-FID. Hasil Analisis tersebut dapat digunakan untuk menentukan yieldbiodiesel dari proses transesterifikasi. Hasil penelitian menunjukkan bahwa yield biodiesel meningkatsecara perlahan dengan meningkatnya waktu reaksi pada keadaan subkritis (210-230oC. Namun

  12. KARAKTERISTIK DAN PENDEKATAN KINETIKA GLOBAL PADA PIROLISIS LAMBAT SAMPAH KOTA TERSELEKSI

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    Dwi Aries Himawanto

    2012-04-01

    dipertahankan pada 400°C  selama 30 menit. Untuk menjamin terjadinya proses pirolisis, maka dialirkan nitrogen dengan laju 100 ml/menit ke dalam reaktor. Hasil penelitian menunjukkan bahwa sampah bambu dan sampah daun pisang dapat dikategorikan kedalam bahan organik dengan kestabila rendah. Sampah bahan pengemas cenderung masuk dalam katagori bahan campuran polimer, sedangkan sampah styrofoam dapat dikategorikan ke dalam material plastik. Hasil penelitian juga menunjukkan bahwa metode kinetika global dapat digunakan untuk memprediksi nilai energi aktivasi dari komponen tunggal sampah kota.

  13. THE PROPERTIES OF CHARCOAL FROM THE BLACK LIQUOR OF THE SODA PULPING OF RICE STRAW

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    Nyoman Jaya Wistara

    2013-11-01

    Full Text Available The main goal of the present works was to determine chemical changes, thermal decomposition, and the content of moisture, ash, volatile, fixed carbon and calorific value of soda pulping black liquor of the rice straw. Neutralized black liquor was dried to a moisture content of 10% and then pyrolized at 106oC-750oC. It was found that calorific value, fixed carbon, volatile mater, and moisture content were in the range of 2782-4716 cal/g, 49.2-81.6%, 15.5-47.5%, and 0.2-3.5%, respectively. Ash content was not influenced by the temperature of pyrolysis and was thought to depend on its initial silicate content. The weight loss of pulp was higher than that of black liquor. Extreme weight loss has been found in the temperature of 200-400oC. Noticeable functional groups changes were found with the increasing temperature of pyrolysis. Hydroxyl group completely disappeared at 300oC and above. Carbonyl related groups were also disappeared at 300-500oC, but it was reformed at 650 and 750oC. It might be brought about by the deformation of chemical bonding of oxygen ring in lignin structures. SIFAT-SIFAT ARANG LINDI HITAM DARI PEMASAKAN JERAMI DENGAN LARUTAN SODA API. Penelitian ini bertujuan untuk menentukan perubahan sifat kimia, dekomposisi termal dan kadar air, abu, zat terbang, karbon terikat serta nilai kalor arang lindi hitam pemasakan soda jerami padi. Dalam penelitian ini, lindi hitam netral dikeringkan (kadar air 10%, kemudian dipirolisis pada selang suhu 100-750oC di dalam reaktor berpengatur suhu. Hasil penelitian menunjukkan bahwa nilai kalor, karbon terikat, zat terbang dan kadar air masing-masing berselangdari 2782-4716 cal/g, 49,2-81,6%, 15,5-47,5%, dan 0,2-3,5%. Kadar abu tidak dipengaruhi oleh suhu pirolisis dan diduga bergantung pada kadar silika bahan bakunya. Nilai kalor meningkat dengan meningkatnya kadar karbon terikat. Perilaku kehilangan berat arang dari lindi hitam berbeda dengan perilaku kehilangan berat pulp jerami. Kehilangan

  14. Calculated activities of some isotopes in the RA reactor highly enriched fuel significant for possible environmental contamination - Operational report; Radni izvestaj - Proracun aktivnosti nekih izotopa u visokoobogacenom uranskom gorivu reaktora RA, znacajnih sa gledista moguce kontaminacije okoline

    Energy Technology Data Exchange (ETDEWEB)

    Bulovic, V; Martinc, R; Cupac, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-12-15

    This report contains calculation basis and obtained results of activities for three groups of isotopes in the RA reactor 80% enriched fuel element. The following isotopes are included: 1) {sup 85m}Kr, {sup 87}Kr, {sup 88}Kr, {sup 131}J, {sup 132}J, {sup 133}J, {sup 134}J, {sup 135}J, {sup 133}Xe, {sup 138}Xe i {sup 138}Cs, 2) {sup 89}Sr, {sup 90}Sr, {sup 91}Sr, {sup 92}Sr, {sup 95}Zr, {sup 97}Zr, {sup 103}Ru, {sup 105}Ru, {sup 106}Ru, {sup 129m}Te, {sup 134}Cs, {sup 137}Cs, {sup 140}Ba, {sup 144}Ce, kao i 3) {sup 238}Pu, {sup 239}Pu i {sup 240}Pu. It was estimated that the fuel is exposed to mean neutron flux. The periodicity of reactor operation is taken into account. Calculation results are given dependent on the time of exposure. These results are to be used as source data for Ra reactor safety analyses. [Serbo-Croat] Izlozene su osnove i prikazani su rezultati izvedenog proracuna aktivnosti tri grupe izotopa u gorivnom elementu reaktora RA sa 80% obogacenim uranom - 235. Obuhvaceni su: 1) {sup 85m}Kr, {sup 87}Kr, {sup 88}Kr, {sup 131}J, {sup 132}J, {sup 133}J, {sup 134}J, {sup 135}J, {sup 133}Xe, {sup 138}Xe i {sup 138}Cs, zatim, 2) {sup 89}Sr, {sup 90}Sr, {sup 91}Sr, {sup 92}Sr, {sup 95}Zr, {sup 97}Zr, {sup 103}Ru, {sup 105}Ru, {sup 106}Ru, {sup 129m}Te, {sup 134}Cs, {sup 137}Cs, {sup 140}Ba, {sup 144}Ce, kao i 3) {sup 238}Pu, {sup 239}Pu i {sup 240}Pu. Pretpostavljeno je da se gorivo ozracuje na srednjem fluksu neutrona, a periodicnost rada reaktora je uvazavana. Rezultati proracuna, dati u numerickom obliku, sistematizovani su kao funkcija toka vremena ozracivanja goriva. Ovi rezultati bice korisceni kao izvorni podaci kod izrade sigurnosnih analiza za reaktor RA (author)

  15. FRM-II research neutron source commissioned; Eroeffnung der Forschungs-Neutronenquelle Heinz Maier-Leibnitz - FRM-II

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2004-07-01

    . Thomas Goppel (Bayerischer Staatsminister fuer Wissenschaft, Forschung und Kunst), Professor Wolfgang A. Herrmann (Praesident der Technischen Universitaet Muenchen), Hannelore Gabor (2. Buergermeisterin der Standortgemeinde Garching) und Professor Winfried Petry (Wissenschaftlicher Direktor des FRM-II) unterstrichen die grosse Bedeutung des FRM-II fuer Wissenschaft Forschung, Medizin und Technik. Der FRM-II wird nicht nur die 'Neutronenluecke' schliessen, sondern mit einem wesentlich hoeheren Neutronenfluss als beim Vorgaenger-Reaktor, dem legendaeren Atom-Ei, ausserordentlich brillante Arbeitsbedingungen schaffen. International steht nach ersten Konzeptionsansaetzen ab 1980, dem 1. Spatenstich im August 1996 und der Ostern 2003 erteilten 3. Teilgenehmigung ein einmaliges Werkzeug fuer moderne Wissenschaften sowie medizinische und technische Anwendungen zur Verfuegung. (orig.)

  16. PENGUJIAN AKTIVITAS KOMPOSIT Fe2O3-SiO2 SEBAGAI FOTOKATALIS PADA FOTODEGRADASI 4-KLOROFENOL (The Activity Test of Fe2O3-SiO2 Composite As Photocatalyst on 4-Chlorophenol Photodegradation

    Directory of Open Access Journals (Sweden)

    Eko Sri Kunarti

    2009-03-01

    Full Text Available ABSTRAK  Pada penelitian ini telah dilakukan pengujian aktivitas komposit Fe2O3-SiO2 sebagai fotokatalis pada fotodegradasi 4-klorofenol. Penelitian diawali dengan preparasi dan karakterisasi fotokatalis Fe2O3-SiO2. Preparasi dilakukan dengan metode sol-gel pada temperatur kamar menggunakan tetraetil ortosilikat (TEOS dan besi (III nitrat sebagai prekursor diikuti dengan perlakuan termal pada temperature 500 oC. Karakterisasi dilakukan dengan metode spektrometri inframerah, difraksi sinar-X dan spektrometri fluoresensi sinar-X. Uji aktivitas komposit untuk fotodegradasi 4-klorofenol dilakukan dalam reaktor tertutup yang dilengkapi dengan lampu UV. Pada uji ini telah dipelajari pengaruh waktu penyinaran dan pH larutan terhadap efektivitas fotodegradasi 4-klorofenol. Hasil penelitian menunjukkan bahwa komposit Fe2O3-SiO2 dapat dipreparasi dengan metode sol-gel pada temperatur kamar diikuti perlakuan termal. Komposit Fe2O3-SiO2 dapat meningkatkan efektivitas fotodegradasi 4-klorofenol dari 11,86 % menjadi 55,38 %. Efektivitas fotodegradasi 4- klorofenol dipengaruhi waktu penyinaran dan pH larutan yang semakin lama waktu penyinaran efektifitas fotodegradasi semakin tinggi, namun waktu penyinaran yang lebih lama dari 4 jam dapat menurunkan efektivitasnya. pH larutan memberikan pengaruh yang berbeda-beda pada efektivitas fotodegradasi 4-klorofenol.   ABSTRACT The activity test of Fe2O3-SiO2 composite as photocatalyst on 4-chlorophenol photodegradation has been studied. The research was initiated by preparation of Fe2O3-SiO2 photocatalyst and followed by characterization. The preparation was conducted by sol-gel method at room temperature using tetraethylorthosilicate (TEOS and iron (III nitrate as precursors followed by thermal treatment at a temperature of 500oC. The characterizations were performed by X-ray Diffraction (XRD, Infrared and X-ray Fluorescence Spectrophotometry. The photocatalytic activity test of composites for 4 chlorophenol

  17. Technical realization of the VISA-2 Project, contract: 2.01/ I phase, Volume No. I; Tehnicka realizacija projekta VISA-2, ugovor: 2.01/I faza, Album br. I

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-12-15

    , radijacionog ostecenja materijala. Projekat je podeljen u tri faze. Prva faza obuhvata radove koji se odnose na rekonstrukciju tehnoloskih kanala reaktora RA za potrebe projekta VISA-2. Druga faza obuhvata radove koji se odnose na merne sisteme kojima ce se meriti temperatura ulazne i izlazne teske vode u 5 kanala VISA-2, kao i temperature uzoraka (55 termoparova) kao i radove na ispitivanju kanala i kapsula pre i posle ugradnje u reaktor. Treca faza obuhvata radove koji se odnose na probleme transporta aktivnih tehnoloskih kanala i kapsula VISA-2, probleme njihovog secenja i odvajanje od kanala, kao i probleme pakovanja i transporta ozracenih uzoraka od Vince do Sacley-a. Ovaj album sadrzi sve dokumente relevantne za izvrsenje zadatka ukljucujuci ugovore, neophodne preliminarne proracune kao i analizu sigurnosti.

  18. THERMAL NEUTRON FLUX MAPPING ON A TARGET CAPSULE AT RABBIT FACILITY OF RSG-GAS REACTOR FOR USE IN k0-INAA

    Directory of Open Access Journals (Sweden)

    Sutisna Sutisna

    2015-03-01

    Full Text Available Instrumental neutron activation analysis based on the k0 method (k0-INAA requires the availability of the accurate reactor parameter data, in particular a thermal neutron flux that interact with a targets inside the target capsule. This research aims to determine and map the thermal neutron flux inside the capsule and irradiation channels used for the elemental quantification using the k0-AANI. Mapping of the thermal neutron flux (фth on two type of irradiation capsule have been done for RS01 and RS02 facilities of RSG-GAS reactor. Thermal neutron flux determined using Al-0,1%Au alloy through 197Au(n,g 198Au nuclear reaction, while the flux mapping done using statistics R. Thermal neutron flux are calculated using k0-IAEA software provided by IAEA. The results showed the average thermal neutron flux is (5.6±0.3×10+13 n.cm-2.s-1; (5.6±0.4×10+13 n.cm-2.s-1; (5.2±0.4×10+13 n.cm-2.s-1 and (5.3±0.4×10+13 n.cm-2.s-1 for Polyethylene capsule of 1st , 2nd, 3rd and 4th layer respectively. In the case of Aluminum capsule, the thermal neutron flux was lower compared to that on Polyethylene capsule. There were (3.0±0.2×10+13 n.cm-2.s-1; (2.8±0.1×10+13 n.cm-2.s-1; (3.2±0.3×10+13 n.cm-2.s-1 for 1st, 2nd and 3rd layers respectively. For each layer in the capsule, the thermal neutron flux is not uniform and it was no degradation flux in the axial direction, both for polyethylene and aluminum capsules. Contour map of eight layer on polyethylene capsule and six layers on aluminum capsule for RS01 and RS02 irradiation channels had a similar pattern with a small diversity for all type of the irradiation capsule. Keywords: thermal neutron, flux, capsule, NAA   Analisis aktivasi neutron instrumental berbasis metode k0 (k0-AANI memerlukan ketersediaan data parameter reaktor yang akurat, khususnya data fluks neutron termal yang berinteraksi dengan inti sasaran di dalam kapsul target. Penelitian ini bertujuan menentukan dan memetakan fluks neutron termal

  19. Indium-Gallium Radiation Contour of the IRT Nuclear Reactor; Circuit d'activation d'indium-gallium dans le reacteur nucleaire IRT; Indij-gallievyj radiatsionnyj kontur yadernogo reaktora IRT; Circuito de radiaciones de indio-galio del reactor IRT

    Energy Technology Data Exchange (ETDEWEB)

    Breger, A K; Ryabukin, Y S; Tulkes, S G; Volkov, E N

    1960-07-15

    -industrielles. (author) [Spanish] Basandose en un trabajo teorico ya publicado, se preparo en el reactor IRT un circuito de radiaciones de indio-galio que constituye una nueva fuente de rayos gamma de elevada intensidad. El primer circuito de este tipo ''RK-1'' se construyo para el reactor IRT en el Instituto de Fisica de la Academia de Ciencias de la Republica Socialista Sovietica de Georgia. Este trabajo estudia los puntos siguientes: calculo de la activacion de la desintegracion del conjunto indio-galio; estructura del circuito RK-1 y su disposicion en el tanque del reactor y en la camara activa; dispositivo de admision de las sustancias liquidas y gaseosas en la zona de la irradiacion; transportador de las sustancias solidas sometidas a irradiacion. En el reactor IRT, cuya potencia es de 2 000 kW, la intensidad de irradiacion del circuito es igual a la de una fuente de radiacion gamma equivalente a 20 000 g de radio. En el trabajo se estudian las posibilidades de utilizacion de este circuito con fines semi-industriales y de investigacion. (author) [Russian] Osnovyvayas' na uzhe opublikovannoj teoreticheskoj rabote, byl podgotovlen indij-gallievyj radiatsionnyj kontur yadernogo reaktora IRT, kotoryj yavlyaetsya novym moshchnym istochnikom gamma-oblucheniya . Pervyj kontur ehtogo tipa RK-1 byl podgotovlen na reaktore IRT v Institute fiziki Akademii nauk Gruzinskoj SSR. V doklade dayutsya raschety aktivizatsij dlya indij-gallievogo splava, strukturnye kompanovki RK-1 i ikh raspolozhenie v bake reaktora i goryachej kamere, ustrojstvo podachi zhidkikh i gazoobraznykh veshchestv v zonu oblucheniya i konvejer dlya tverdykh veshchestv, kotorye podlezhat oblucheniyu. V reaktore IRT moshchnost'yu 2000 kW moshchnost' oblucheniya kontura ehkvivalentna moshchnosti oblucheniya gamma-izluchatelya, obladayushchego aktivnost'yu v 20000 g ehkv. radiya. V doklade obsuzhdayutsya perspektivy ispol'zovaniya indij-gallievogo radiatsionnogo kontura dlya issledovatel'skikh i polupromyshlennykh tselej

  20. DESAIN AWAL TURBIN UAP TIPE AKSIAL UNTUK KONSEP RGTT30 BERPENDINGIN HELIUM

    Directory of Open Access Journals (Sweden)

    Sri Sudadiyo

    2016-06-01

    Full Text Available ABSTRAK DESAIN AWAL TURBIN UAP TIPE AKSIAL UNTUK KONSEP RGTT30 BERPENDINGIN HELIUM. Konsep reaktor daya nuklir yang dikembangkan merupakan jenis reaktor berpendingin gas dengan temperatur tinggi (RGTT. Gas yang digunakan untuk mendinginkan teras RGTT adalah helium. Konsep RGTT ini dapat menghasilkan daya termal 30 MWth sehingga dinamakan RGTT30. Temperatur helium mampu mencapai 700 °C ketika keluar dari teras RGTT30 dan digunakan untuk memanaskan air di dalam steam generator hingga mencapai temperatur 435 °C. Steam generator dihubungkan dengan turbin uap yang dikopel dengan generator listrik untuk membangkitkan daya 7,27 MWe. Uap yang keluar dari turbin dilewatkan kondensor untuk mencairkan uap menjadi air. Rangkaian komponen dari steam generator, turbin, dan kondensor dinamakan sistem turbin uap. Turbin terdiri dari sudu-sudu yang dimaksudkan untuk mengubah tenaga uap kedalam tenaga mekanis berupa putaran. Efisiensi turbin merupakan parameter yang harus diperhatikan dalam sistem turbin uap ini. Tujuan dari makalah ini adalah untuk mengusulkan sudu tipe aksial dan untuk menganalisa perbaikan efisiensi turbin. Metode yang digunakan yaitu aplikasi prinsip termodinamika yang berhubungan dengan konservasi energi dan massa. Perangkat lunak Cycle-Tempo dipakai untuk mendapatkan parameter termodinamika dan untuk mensimulasikan sistem turbin uap berbasis RGTT30. Pertama, dibuat skenario dalam simulasi sistem turbin uap untuk mengetahui efisiensi dan laju aliran massa uap yang diperoleh nilai optimal 87,52 % dan 8,759 kg/s pada putaran 3000 rpm. Kemudian, turbin uap diberi sudu tipe aksial dengan diameter tip 1580 mm dan panjang 150 mm. Hasil yang diperoleh adalah nilai efisiensi turbin uap naik menjadi 88,3 % pada putaran konstan (3000 rpm. Penambahan nilai efisiensi turbin sebesar 0,78 % menunjukkan peningkatan kinerja RGTT30 secara keseluruhan. Kata kunci: Tipe aksial, turbin uap, RGTT30   ABSTRACT PRELIMINARY DESIGN ON STEAM TURBINE OF AXIAL TYPE

  1. PHASE CHANGES ON 4H AND 6H SIC AT HIGH TEMPERATURE OXIDATION

    Directory of Open Access Journals (Sweden)

    Jan Setiawan

    2016-10-01

    -016-4971 card.  Diffraction pattern on 46S also showed lattice parameter, composition and crystallite size changes.  The lattice parameter changes not significant.  For 6S and 46S sam-ples at 1400 oC, the 6H-SiC phase changes into other phases more than 50 % from its original weight percentage. Keywords: silicon carbide, 4H-SiC, 6H-SiC, oxidation, high temperature. ABSTRAK PERUBAHAN FASA 4H DAN 6H SIC YANG TEROKSIDASI PADA TEMPERATUR TINGGI.  Telah dilakukan proses oksidasi pada silikon karbida yang mengadung fasa 6H dan silikon karbida yang mengandung fasa 4H dan 6H.  Silikon karbida merupakan keramik non oksida dengan sifat-sifat unggulnya yang sangat potensial digunakan dalam dunia industri.  Dalam industri nuklir silikon karbida digunakan sebagai bahan struktur kelongsong pada bahan bakar reaktor air ringan light water reactor (LWR dan sebagai pelapis pada kernel bahan bakar reaktor gas temperatur tinggi (RGTT.  Pada studi ini dilakukan simulasi oksidasi silikon karbida pada kernel apabila terjadi kegagalan pada pipa pendingin utamanya. Sampel dibentuk dari serbuk silikon karbida yang di pres hingga berbentuk pelet dengan diameter 12,7 mm dan ketebalan 1.0 mm kemudian dioksidasi pada temperatur 1000 oC, 1200 oC dan 1400 oC selama 1 jam.  Sampel sebelum dan setelah dioksidasi dilakukan penimbangan dan pengujian difraksi sinar-X menggunakan Difraktometer Panalytical Empyrean dengan Cu sebagai sumber sinar-X.  Analisis pola difraksi dilakukan menggunakan aplikasi General Structure Analysis System (GSAS, dengan hasil yang diperoleh adalah perubahan parameter kisi dan kandungan fasa SiC-nya.  Hasil percobaan menunjukkan bahwa semua sampel yang teroksidasi mengalami peningkatan berat.  Oksidasi sampel 6S menyebabkan kenaikan berat tertinggi pada temperatur 1200 oC, sedangkan sampel 46S memiliki berat dengan kecenderungan meningkat seiring dengan meningkatnya temperatur oksidasi.  Analisis pola difraksi sinar-X menunjukkan bahwa fasa domi-nan yang terbentuk pada sampel

  2. Post-Construction Testing of the Elk River, Hallam and Piqua Power Reactor Plants; Essais apres construction des centrales nucleaires d'Elk River, de Hallam et de Piqua; Predehkspluatatsionnoe ispytanie Ehlk-riverskoj, Khehlpemskoj i Pikuaskoj ehnergeticheskikh reaktornykh ustanovok; Ensayos posteriores a la construccion de las centrales nucleoelectricas de Elk River, Hallam y Piqua

    Energy Technology Data Exchange (ETDEWEB)

    Pursel, C. A. [United States Atomic Energy Commission, Argonne, IL (United States)

    1963-10-15

    defectos hallados: Reactor de Elk River. Se descubrieron grietas en parte del revestimiento superficial del recipiente del reactor; ello obligo a efectuar una serie de investigaciones y analisis, asi como ciertas reparaciones y modificaciones del recipiente. La insuficiente capacidad de separacion de vapor obligo a sustituir y modificar algunas piezas metalicas en el interior del recipiente del reactor. Central nucleoelectrica de Hallam. Debido al arrastre de helio, hubo que modificar los circuitos secundarios de sodio. La falla de un tubo del intercambiador de calor intermedio (sodio-sodio) obligo a llevar a cabo una serie de analisis para descubrir su causa y extraer y reparar el intercambiador. Central nucleoelectrica de Piqua. Durante la limpieza de las tuberias con agentes quimicos, se dallaron varias valvulas que fue preciso reparar o sustituir. Las fugas en el circuito del refrigerante organico y del vapor secundario provocaron demoras repetidas. Una vez concluidas las reparaciones e introducidas las modificaciones necesarias, se comprobo que las caracteristicas de rendimiento reales de cada uno de los tres reactores se ajustaban estrictamente a las previstas en el proyecto. (author) [Russian] Fakticheskij opyt, nakoplennyj vo vremya predehkspluatatsionnykh ispytanij trekh yadernykh ehnergeticheskikh ustanovok, postroennykh po demonstratsionnoj programme ehnergeticheskikh reaktorov Komissii po atomnoj ehnergii Soedinennykh Shtatov, pozvolyaet sdelat' nekotorye obobshcheniya v otnoshenii ehtoj fazy stroitel'stva i ehkspluatatsii ustanovok. Tri ustanovki, a imenno Ehlk-riverskij reaktor (ERR), Khehllemskaya yadernaya ehnergeticheskaya ustanovka (HNPF) i Pikuaskaya yadernaya ehnergeticheskaya ustanovka (PNPF), predstavlyayut tri razlichnykh tipa reaktorov: reaktor s kipyashej vodoj s estestvennoj tsirkulyatsiej, natrievo-grafitovyj reaktor i reaktor s organicheskim teplonositelem i zamedlitelem sootvetstvenno. Period predehkspluatatsionnykh ispytanij okhvatyvaet vremya

  3. The Measurement of Reactivity In Multiregion Subcritical Systems by the Pulsed Neutron Technique; Mesure de la Reactivite dans les Systemes Sous-Critiques a Plusieurs Regions par la Methode des Neutrons Pulses; Izmerenie reaktivnosti v mnogozonnykh podkriticheskikh sistemakh metodom impul'snykh nejtronov; Mediciones de la Reactividad en Sistemas Subcriticos de Varias Regiones Mediante la Tecnica de los Neutrones Pulsados

    Energy Technology Data Exchange (ETDEWEB)

    Sherwin, J.; Leng, J. H. [United Kingdom Atomic Energy Authority, Windscale Works, Cumberland (United Kingdom)

    1965-10-15

    se obtuvo en el reactor avanzado de Windscale, refigerado con gas, cuyo funcionamiento se interrumpio deliberadamente por un envenenamiento uniforme. En el segundo caso se hizo variar la reactividad del cuerpo del reactor 'HERO' de potencia nula modificando el radio de la zona cargada. Todas las mediciones de los impulsos ncuironicos concuerdan satisfactoriamente con las obtenidas por otros metodos mas clasicos de determinacion de la reactividad. Por ultimo, los autores examinan las tecnicas experimentales y las dificultades encontradas en los sistemas moderados con grafito. (author) [Russian] Mgnovennyj raspad termalizovannoj vspyshki nejtronov v mnogojeonnoj podkriticheskoj sisteme izuchalsja s pomoshh'ju dvuhgruppovoj teo- rii diffuzii. Pokazyvaetsja, chto mozhno ustanovit' svjaz' mezhdu postojannoj mgnovennogo raspada osnovnogo sostojanija i kojefficientom jeffektivnogo razmnozhenija sistemy v ramkah dvuh opredelennyh parametrov dlja celej publikuemogo doklada v kachestve kojefficienta chuvstvitel'nosti reaktora i popravki na mgnovennyj raspad, kotorye zavisjat v bol'shoj stepeni ot prostranstvennogo razmeshhenija potokov v sisteme. Pri odnorodnoj sisteme bez otrazhatelja kojefficient chuvstvitel'nosti mozhno ustanovit' pri opredelenii srednego perioda zhizni nejtrona v sisteme, dlja mnogozonnoj sistemy takoj kojefficient predstavljaet soboj sochetanie periodov zhizni v kazhdoj zone, uslozhnennyh in- tegralami vozmushhenija. Vtoroj parametr, popravka na raspad mozhet che imet' fizicheskogo smysla v svjazi s tem, chto ona pojavljaetsja pri popytke ustanovit' svjaz' mezhdu dvumja so- otvetstvujushhimi masshtabami reaktivnosti, to-est' masshtabom pri opredelenii kotorogo ispol'zuetsja postojannaja mgnovennogo raspada, i masshtabom, vyvodimym s ispol'zovaniem kojefficienta jeffektivnogo razmnozhenija. Izuchajutsja svojstva jetih parametrov s uporom na reaktor s obogashhennym uranom i grafitovym zamedlitelem, sostojashhij iz odnorodnoj aktivnoj zony i otrazhatelja, i

  4. Notes on the Start-Up of the Latina Power Station; Notes concernant le demarrage de la centrale nucleaire de Latina; Zapusk ehlektrostantsii Latina; Notas sobre la puesta en marcha de la central electrica de Latina

    Energy Technology Data Exchange (ETDEWEB)

    Calabria, G.; Gualtieri, G. [AGIP Nucleare, Milano (Italy)

    1963-10-15

    desarrollo de los ensayos finales, sobre la carga del combustible y el orden en que se ejecutaron las operaciones de puesta en marcha y sobre las determinaciones y maniobras de regulacion posteriores al estado critico. Se resenan asimismo las operaciones iniciales de generacion y conexion de la central con la red electrica. Por ultimo, se mencionan los problemas de organizacion derivados de la explotacion de la central, incluyendo la preparacion y formacion profesional del personal y las medidas de seguridad adoptadas. (author) [Russian] Privoditsya informatsiya o zapuske pervoj ital'yanskoj atomnoj ehlektrostantsii Latina moshchnost'yu 200 mgvt. Reaktor rabotaet na prirodnom urane s grafitovym zamedlitelem i gazovym okhlazhdeniem. Posle kratkogo opisaniya osnovnykh kharakteristik ehlektrostantsii privodyatsya podrobnye dannye otnositel'no provedeniya zaklyuchitel'nykh ispytanij stantsii'', toplivnoj zagruzki i poryadka ehkspluatatsii, izmerenij i upravleniya s dovedeniem do kriticheskogo sostoyaniya. Daetsya ob''yasnenie raboty po zapusku i vklyucheniyu stantsii v ehlektroset'. Izlagayutsya takzhe problemy, svyazannye s ehkspluatatsiej stantsii, v tom chisle podgotovka personela i ego kvalifikatsiya, mery po bezopasnosti. (author)

  5. OPTIMASI LAJU ALIR MASSA DALAM PURIFIKASI PENDINGIN RGTT200K UNTUK PROSES KONVERSI KARBONMONOKSIDA

    Directory of Open Access Journals (Sweden)

    Sumijanto Sumijanto

    2016-03-01

    Full Text Available ABSTRAK OPTIMASI LAJU ALIR MASSA DALAM PURIFIKASI PENDINGIN RGTT200K UNTUK PROSES KONVERSI KARBONMONOKSIDA. Karbonmonoksida adalah spesi yang sulit dipisahkan dari helium pendingin reaktor karena mempunyai ukuran molekul relatif kecil sehingga diperlukan proses konversi menjadi karbondioksida. Laju konversi karbonmonoksida dalam sistem purifikasi dipengaruhi oleh beberapa parameter diantaranya konsentrasi, temperatur dan laju alir massa. Dalam penelitian ini dilakukan optimasi laju alir massa dalam purifikasi pendingin RGTT200K untuk proses konversi karbonmonoksida. Optimasi dilakukan melalui simulasi proses konversi karbonmonoksida menggunakan perangkat lunak Super Pro Designer. Laju pengurangan spesi reaktan, laju pertumbuhan spesi antara dan spesi produk dalam kesetimbangan reaksi konversi dianalisis untuk memperoleh optimasi laju alir massa purifikasi terhadap proses konversi karbonmonoksida. Tujuan penelitian ini adalah menyediakan data laju alir massa purifikasi untuk pembuatan dasar desain sistem purifikasi helium pendingin RGTT200K. Hasil analisis menunjukkan bahwa pada laju alir massa 0,6 kg/detik proses konversi belum optimal, pada laju alir massa 1,2 kg/detik mencapai optimal dan pada laju alir 3,6 kg/detik s/d 12,0 kg/detik tidak efektif. Untuk memdukung dasar desain sistem purifikasi helium pendingin RGTT200K maka laju alir massa purifikasi untuk proses konversi karbonmonoksida digunakan laju alir massa 1,2 kg/detik. Kata kunci: Karbonmonoksida, konversi, purifikasi, laju alir massa, RGTT200K.   ABSTRACT OPTIMIZATION OF MASS FLOW RATE IN RGTT200K COOLANT PURIFICATION FOR CARBONMONOXIDE CONVERSION PROCESS. Carbonmonoxide is a species that is difficult to be separated from the reactor coolant helium because it has a relatively small molecular size.  So it needs a process of conversion from carbonmonoxide to carbondioxide. The rate of conversion of carbonmonoxide in the purification system is influenced by several parameters

  6. PENINGKATAN KUALITAS DAN PROSES PEMBUATAN BIODIESEL DARI BLENDING MINYAK KELAPA SAWIT (PALM OIL DAN MINYAK KELAPA (COCONUT OIL DAN BANTUAN GELOMBANG ULTRASONIK

    Directory of Open Access Journals (Sweden)

    Hantoro Satriadi

    2015-01-01

    Full Text Available Keterbatasan solar sebagai sumber energi bahan bakunya tidak dapat diperbaharui menuntut adanya bahan baku alternatif yang dapat diperbaharui dan ramah lingkungan untuk pembuatan biodiesel. Reaksi utama produksi biodiesel adalah esterifikasi dan transestirifikasi yang berlangsung lambat dan membutuhkan banyak katalis dan alkohol. Reaksi yang terjadi belum sempurna dan belum memenuhi standar SNI dan ASTM. Untuk memperbaiki mutu biodiesel serta menghasilkan yield maksimal, maka dilakukan blending bahan baku antara minyak kelapa sawit dan minyak kelapa dan dengan bantuan gelombang ultrasonic. Penelitian ini bertujuan untuk mempelajari pengaruh variabel perbandingan volume minyak kelapa sawit dan minyak kelapa, perbandingan volume methanolminyak, dan persentase berat katalis terhadap minyak terhadap hasil atau yield biodiesel. Alat utama yang digunakan adalah reaktor yang dilengkapi pembangkit gelombang ultrasonic dengan temperature 60 oC, tekanan 1 atm, volume 3 liter, dan frekuensi 28 kHz. Variabel proses pada penelitian ini adalah perbandingan volume minyak sawit dan kelapa 2:1, 3:1, dan 4:1, pebandingan volume metanol-minyak 0,2:1, 0,25:1, dan 0,3:1, dan persentase berat katalis KOH terhadap minyak 0,3%, 0,5%, dan 0,7%. Hasil penelitian didapat konversi tertinggi dicapai pada variabel perbandingan volume minyak sawit dan kelapa 3:1, perbandingan volume metanol/minyak 0,25:1, dan persentase berat katalis terhadap minyak dengan yield 97,26%.[A Improvement of Quality and Process for Biodiesel Production from Palm Oil and Coconut Oil Blends with Ultrasound Assisted] Limitations of solar energy as a source of raw material cannot be renewed demands for alternative raw materials that are renewable and environmentally friendly for the manufacture of biodiesel. The main production of biodiesel reaction is esterification and transestirifikasi which runs slow and requires a lot of alcohol and a catalyst. Reactions that happen yet perfect, and has not met

  7. KARAKTERISASI DAN AKTIVITAS KATALITIK BERBAGAI VARIASI KOMPOSISI KATALIS Ni DAN ZnBr2 DALAM Γ-Al2O3 UNTUK ISOMERISASI DAN HIDROGENASI (R-(+-SITRONELAL

    Directory of Open Access Journals (Sweden)

    ED Iftitah

    2014-06-01

    Full Text Available Pengaruh sifat dan karakter berbagai variasi komposisi katalis Ni dan ZnBr2 yang terimpregnasi dalam γ-Al2O3 terhadap aktivitas dan selektivitasnya untuk reaksi isomerisasi dan hidrogenasi (R-(+-Sitronelal telah dilakukan. Dalam penelitian ini, terdapat tiga jenis variasi komposisi Ni dan ZnBr2 dalam γ-Al2O3, yaitu: A1=Ni/ZnBr2/γ-Al2O3 (3:2, A2=Ni/ZnBr2/γ-Al2O3 (1:1 dan A3=Ni/ZnBr2/γ-Al2O3 (2:3. Katalis dikarakterisasi menggunakan X-ray diffraction (XRD, Brunauer-Emmett-Teller (BET surface area, dan SEM-EELS. Luas area permukaan spesifik dan porositasnya ditentukan berdasarkan adsorption-desorption gas nitrogen pada 77 K. Distribusi dan volume pori ditentukan dengan desorption isotherm pada P/Po ≥ 0,3. Hasil yang didapat menunjukkan bahwa terdapat hubungan antara karakter dan sifat katalis dengan aktivitas katalitiknya terhadap produk isomerisasi dan hidrogenasi (R-(+-Sitronelal. Uji aktivitas dilakukan dalam sebuah reaktor mini dengan 0,5 g katalis dan 3 mL (R-(+-Sitronelal menggunakan atmosfir gas N2 dan/atau H2 dalam waktu 5 dan 24 jam masing-masing pada suhu 90 dan 120 °C. Komposisi katalis, pemilihan jenis atmosfir gas dan suhu sangat berpengaruh terhadap aktivitas dan selektivitas pembentukan produk isomerisasi dan hidrogenasi. Konversi (R-(+-Sitronelal tertinggi ditunjukkan oleh katalis A3=Ni/ZnBr2/γ-Al2O3 (2:3 dengan kondisi reaksi selama 5 jam (4 jam N2 + 1 jam H2 pada suhu 90 °C dan 24 jam (4 jam N2 + 20 jam H2 pada suhu 120 °C. The influence of catalyst properties and characteristics of Ni and ZnBr2 catalysts impregnated in γ-Al2O3 on the activity and selectivity of (R-(+-Citronellal isomerisation and hydrogenation has been done. In this study, there were three sets of Ni and ZnBr2 in γ-Al2O3 with various composition, they were A1=Ni/ZnBr2/γ-Al2O3 (3:2, A2=Ni/ZnBr2/γ-Al2O3 (1:1, A3=Ni/ZnBr2/γ-Al2O3 (2:3. The catalysts were characterized using X-ray diffraction (XRD, Brunauer-Emmett-Teller (BET surface area and SEM

  8. technical guidelines for the design and construction of the next generation of nuclear power plants with pressurized water reactors

    International Nuclear Information System (INIS)

    2009-01-01

    These technical guidelines present the opinion of the French 'Groupe Permanent charge des Reacteurs nucleaires' (GPR) concerning the safety philosophy and approach as well as the general safety requirements to be applied for the design and construction of the next generation of nuclear power plants of the PWR (pressurized water reactor) type, assuming the construction of the first units of this generation would start at the beginning of the 21. century. These technical guidelines are based on common work of the French Institut de Protection et de Surete Nucleaire (IPSN) and of the German Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). Moreover, these technical guidelines were extensively discussed with members of the German Reaktor Sicherheitskommission (RSK) until the end of 1998 and further with German experts. The context of these technical guidelines must be clearly understood. Faced with the current situation of nuclear energy in the world, the various nuclear steam supply system designers are developing new products, all of them claiming their intention of obtaining a higher safety level, by various ways. GPR believes that, for the operation of a new series of nuclear power plants at the beginning of the next century, the adequate way is to derive the design of these plants in an 'evolutionary' way from the design of existing plants, taking into account the operating experience and the in-depth studies conducted for such plants. Nevertheless, introduction of innovative features must also be considered in the frame of the design of the new generation of plants, especially in preventing and mitigating severe accidents. GPR underlines here that a significant improvement of the safety of the next generation of nuclear power plants at the design stage is necessary, compared to existing plants. If the search for improvement is a permanent concern in the field of safety, the necessity of a significant step at the design stage clearly derives from better

  9. KARAKTERISTIK MIKROSTRUKTUR DAN FASA PADUAN Zr- 0,3%Nb-0,5%Fe-0,5%Cr PASCA PERLAKUAN PANAS DAN PENGEROLAN DINGIN

    Directory of Open Access Journals (Sweden)

    Sungkono Sungkono

    2015-07-01

    Full Text Available KARAKTERISTIK MIKROSTRUKTUR DAN FASA PADUAN Zr-0,3%Nb-0,5%Fe-0,5%Cr PASCA PERLAKUAN PANAS DAN PENGEROLAN DINGIN. Logam paduan Zr-Nb-Fe-Cr dikembangkan sebagai material kelongsong elemen bakar dengan fraksi bakar tinggi untuk reaktor daya maju. Dalam penelitian ini telah dibuat paduan Zr-0,3%Nb-0,5%Fe-0,5%Cr yang mendapat perlakuan panas pada temperatur 650 dan 750°C dengan waktu penahanan 1–2 jam. Tujuan penelitian adalah mendapatkan karakter paduan Zr-0,3%Nb-0,5%Fe-0,5%Cr pasca perlakuan panas dan pengerolan dingin yaitu mikrostruktur, struktur kristal dan fasa-fasa yang ada dalam paduan. Hasil penelitian menunjukkan bahwa paduan Zr-0,3%Nb-0,5%Fe-0,5%Cr pasca perlakuan panas (650ºC, 1-2 jam mempunyai struktur butir ekuiaksial dengan ukuran butir bertambah besar seiring dengan bertambahnya waktu penahanan. Sementara itu, pasca perlakuan panas (750ºC, 1-2 jam terjadi perubahan mikrostruktur paduan dari butir ekuiaksial dan kolumnar menjadi butir ekuiaksial lebih besar. Paduan Zr-0,3%Nb-0,5%Fe-0,5%Cr pasca perlakuan panas (650°C, 1 jam dan (750°C, 1 jam tidak dapat dirol dingin dengan reduksi tebal 5 – 10%, sedangkan pasca perlakuan panas (650ºC, 2 jam dan (750°C, 1.5-2 jam mampu menerima deformasi dingin dengan reduksi ketebalan 5-10% tanpa mengalami keretakan. Senyawa Zr2Fe, ZrCr2 dan FeCr teridentifikai dari hasil uji kristalografi paduan Zr-0,3%Nb-0,5%Fe-0,5%Cr.   MICROSTRUCTURE AND PHASE CHARACTERISTICSOF Zr-0.3%Nb-0.5%Fe-0.5%Cr ALLOY POST HEAT TREATMENT AND COLD ROLLING. Zr-Nb-Fe-Cr alloys was developed as fuel elements cladding with high burn up for advanced power reactors. In this research has been made of Zr-0.3% Nb-0.5% Fe-0.5% Cr alloy were heat treated with varying temperatures at650 and 750°C for 1 until 2 hours. The objectives of this research was to obtain the character of Zr-0.3% Nb-0.5% Fe-0.5% Cr alloy post heat treatment and cold rolling, microstructure nomenclature, crystal structure and phases that presents in the

  10. Development of the neutron-transport code TransRay and studies on the two- and three-dimensional calculation of effective group cross sections; Entwicklung des Neutronentransportcodes TransRay und Untersuchungen zur zwei- und dreidimensionalen Berechnung effektiver Gruppenwirkungsquerschnitte

    Energy Technology Data Exchange (ETDEWEB)

    Beckert, C.

    2007-12-19

    Reaktorkernrechnungen mit 2D-Zellcodes. Ziel dieser Arbeit war es, einen 3D-Zellcode zu entwickeln, mit diesem Code 3D-Effekte zu untersuchen und die Notwendigkeit einer 3D-Datenaufbereitung der Neutronenwirkungsquerschnitte zu bewerten. Zur Berechnung des Neutronentransports wurde die Methode der Erststosswahrscheinlichkeiten, die mit der Ray-Tracing-Methode berechnet werden, gewaehlt. Die mathematischen Algorithmen wurden in den 2D/3D-Zellcode TransRay umgesetzt. Fuer den Geometrieteil des Programms wurde das Geometriemodul eines Monte-Carlo-Codes genutzt. Das Ray-Tracing in 3D wurde auf Grund der hohen Rechenzeiten parallelisiert. Das Programm TransRay wurde an 2D-Testaufgaben verifiziert. Fuer einen Druckwasser-Referenzreaktor wurden folgende 3D-Probleme untersucht: Ein teilweise eingetauchter Regelstab und Void (Vakuum oder Dampf) um einen Brennstab als Modell einer Dampfblase. Alle Probleme wurden zum Vergleich auch mit den Programmen HELIOS (2D) und MCNP (3D) nachgerechnet. Die Abhaengigkeit des Multiplikationsfaktors und der gemittelten Zweigruppenquerschnitte von der Eintauchtiefe des Regelstabes bzw. von der Hoehe der Dampfblase wurden untersucht. Die 3D berechneten Zweigruppenquerschnitte wurden mit drei ueblichen Naeherungen verglichen: Lineare Interpolation, Interpolation mit Flusswichtung und Homogenisierung. Am 3D-Problem des Regelstabes zeigte sich, dass die Interpolation mit Flusswichtung eine gute Naeherung ist. Demnach ist hier eine 3D-Datenaufbereitung nicht notwendig. Beim Testfall des einzelnen Brennstabs, der von Void umgeben ist, erwiesen sich die drei Naeherungen fuer die Zweigruppenquerschnitte als unzureichend. Demnach ist eine 3D-Datenaufbereitung notwendig. Die einzelne Brennstabzelle mit Void kann als der Grenzfall eines Reaktors angesehen werden, in dem sich eine Phasengrenzflaeche herausgebildet hat. (orig.)

  11. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report; Qualifizierung des Kernmodells DYN3D im Komplex mit dem Stoerfallcode ATHLET als fortgeschrittenes Werkzeug fuer die Stoerfallanalyse von WWER-Reaktoren. T. 1. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    Naturumlaufverhalten an thermohydraulischen Versuchsanlagen und der Loesung von Benchmarkaufgaben zu reaktivitaetsinduzierten Transienten, - Akquisition und Aufbereitung von Messdaten zu Transienten aus Kernkraftwerken, Validierung von ATHLET-DYN3D anhand der Nachrechnung eines Stoerfalls mit verzoegerter Schnellabschaltung und einer Pumpentransiente in WWER-Reaktoren, - eine ergaenzende Weiterentwicklung von DYN3D durch Erweiterung der neutronen-physikalischen Datenbasis, Einbau eines verbesserten Modells fuer die Kuehlmittelvermischung, Beruecksichtigung der Nachzerfallswaerme, Berechnung von Xenon-Oszillationen, Analyse von Frischdampfleckszenarien fuer eine WWER-440-Anlage mit Annahme des Versagens verschiedener Sicherheitseinrichtungen, Untersuchung verschiedener Modelloptionen. Die Analyse ergab eine moegliche Rekritikalitaet des abgeschalteten Reaktors bei realistischer Modellierung der Kuehlmittelvermischung im Ringspalt und unteren Plenum. Mit der Anwendung des Programmpakets ATHLET-DYN3D in Tschechien, Bulgarien und der Ukraine wurde bereits begonnen. Weiterfuehrende Arbeiten beinhalten die Verifikation von ATHLET-DYN3D mit einer DYN3D-Version fuer die quadratische Brennelementgeometrie westlicher Druckwasserreaktoren. (orig.)

  12. SINTESIS DIMETIL ASETAL SITRONELAL DENGAN KATALIS GAS HCL

    Directory of Open Access Journals (Sweden)

    E Cahyono

    2014-06-01

    Full Text Available Perlindungan gugus aldehida melalui pembentukan asetal umumnya dilakukan dengan metanol atau etanol terkatalis asam. Sitronelal memiliki gugus aldehida dan gugus alkena. Dalam lingkungan asam, sitronelal mudah mengalami siklisasi membentuk isopulegol dan isomernya. Penelitian ini bertujuan untuk mensintesis dimetil asetal sitronelal dengan katalis gas HCl. Penggunaan gas HCl secara terbatas dimaksudkan untuk menghindari siklisasi sitronelal. Dalam suatu reaktor, 10 mL sitronelal ditambah dengan 20 mL metanol absolut dan 2 g CaCl2 pada labu leher tiga. Gas HCl dialirkan dengan laju alir 12 mL/menit pada temperatur ruang hingga pH campuran menjadi 2-3. Kemudian dilakukan pengadukan pada 30°C selama 48 jam dan diambil sampel pada durasi reaksi 12, 24 dan 48 jam. Dalam penelitian ini dilakukan variasi temperatur dan jumlah CaCl2. Dimetil asetal sitronelal yang dihasilkan diisolasi dengan distilasi fraksinasi pengurangan tekanan dan diuji strukturnya dengan GC-MS, IR dan 1H-NMR. Peningkatan temperatur dan jumlah CaCl2 meningkatkan konversi sitronelal. Setelah 48 jam kuantitas dimetil asetal sitronelal mencapai 48,65%. Distilasi fraksinasi pengurangan tekanan (5 cmHg terbukti meningkatkan kemurnian dimetil asetal sitronelal menjadi 86,39% terhadap produk kasarnya. Elusidasi struktur dengan spektrofotometer infra merah (IR dan resonansi magnetik inti (1H-NMR pada hasil distilasi fraksinasi membuktikan adanya struktur asetal sitronelal.  Aldehyde group protection through acetal formation is generally performed by acid catalyzed methanol or ethanol. Citronellal that has aldehyde and alkene groups. In acidic environment, it is prone to do cyclization to form isopulegol and its isomers. This study aims to synthesize dimethyl acetal of citronellal with HCl gas catalysts. The limitation of HCl using gas was intended to avoid the citronellal cyclization. In a reactor, 10 mL citronellal was added with 20 mL of absolute methanol and 2 g CaCl2 in the three

  13. HIDROLISIS SELULOSA MENJADI GLUKOSA DENGAN KATALIS HETEROGEN ARANG AKTIF TERSULFONASI

    Directory of Open Access Journals (Sweden)

    Didi Dwi Anggoro

    2014-12-01

    . Hasilnya kemudian dicuci dan dikeringkan, dilakukan uji karakteristik dan performance (kinerja katalis berupa kapasitas H+, ukuran pori katalis dengan BET, uji gugus fungsi dengan FTIR, dan uji struktur marfologi katalis dengan SEM. Kinerja katalis diuji dalam reaktor autoclave melalui proses hidrotermal dengan mevariasikan jumlah katalis, dan variasi temperatur. Hasil penelitian menunjukkan untuk uji karakteristik kapasitas H+ sebesar 2,95 mmol/gr, untuk uji BET ukuran pori 29 m2/gr , untuk uji FTIR keberadaan gugus sulfonat terbaca pada vibrasi panjang gelombang 1750 cm-1 dan 1379 cm-1 , pada uji SEM struktur morfologi katalis yang lebih terbuka pada karbon aktif setelah proses sulfonasi. Kinerja katalis konversi tertinggi selulosa menjadi glukosa mencapai 87,2% pada jumlah alang-alang 2 gr, jumlah katalis 2 gr, dan temperatur 170oC selama  8 jam. Kata kunci : selulosa; glukosa; karbon aktif tersulfonasi; alang-alang

  14. The Application of Non-Metallic Core Materials in a High-Temperature Reactor Experiment; Utilisation de materes non metalliques dans le coeur d'un reacteur experimental a haute temperature; Ispol'zovanie nemetallicheskikh materialov dlya aktivnoj zony vysokotemperaturnogo opytnogo reaktora; Empleo de materiales no metalicos en el nucleo de un reactor experimental de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Huddle, R. A.U.; Shepherd, L. R. [Organization for Economic Co-Operation and Development, Dragon Project, Atomic Energy Establishment, Winfrith, Dorset (United Kingdom)

    1963-11-15

    comportamiento de estos materiales en condiciones de funcionamiento normales. Se comunican los resultados de las investigaciones sobre irradiacion, asi como de los trabajos efectuados en los circuitos del reactor. El objetivo principal de este programa es el perfeccionamiento de los reactores de elevada temperatura refrigerados por gas para aplicarlos a la generacion de electricidad en condiciones rentables. (author) [Russian] Proekt vysokotemperaturnogo reaktora (DRAGON)sozdan dlya razrabotki tekhnologii vysokotemperaturnykh reaktorov s gazovym okhladitelem; v nem predusmatrivalos' sooruzhenie i ehkspluatatsiya opytnogo reaktora na 20 mgvt (tepl.). Reaktor, sooruzhenie kotorogo budet vskore zakoncheno, predstavlyaet soboj sistemu, okhlazhdaemuyu geliem; temperatura na vykhode iz aktivnoj zony budet dostigat' 750{sup o}C. V nem budet ispol'zovat'sya U-235 v kachestve goryuchego i torij v kachestve vosproizvodyashchego materiala. Kharakternoj osobennost'yu sistemy yavlyaetsya otsutstvie kakogo-libo metalla v aktivnoj zone. Vvidu togo, chto v reaktore dolzhny razvivat'sya ves'ma vysokie temperatury,' a imenno, 1050{sup o}C na poverkhnosti teplovydelyayushchego ehlementa i do 1500{sup o}C v naibolee sil'no nagrevaemykh tochkakh topliva, dlya ego sooruzheniya ispol'zovany ogneupornye nemetallicheskie materialy. Vse veshchestvo aktivnoj zony sosredotocheno v teplovydelyayushchem ehlemente, blagodarya chemu sootnoshenie mezhdu poverkhnost'yu teploperedachi i ob{sup e}mom aktivnoj zony dostigaet bol'shogo znacheniya, i, sledovatel'no, pozvolyaet dostigat' vysokoj srednej plotnosti ehnergii v sravnitel'no kompaktnoj sisteme. Kazhdyj teplovydelyayushchij ehlement sostoit ieh gruppy grafitovykh trubok, zapolnennykh grafitovymi tabletkami, soderzhashchimi rasshcheplyayushcheesya i vosproizvodyashchee veshchestva v vide karbidov. Gelievyj okhladitel' prokhodit po osi. kazhdogo teplovydelyayushchego sterzhnya i vyvoditsya u ego osnovaniya, okhladitel' zatem napravlyaetsya v ochistitel

  15. PENGARUH TEMPERATUR DAN IRADIASI TERHADAP INTERDIFUSI PARTIKEL BAHAN BAKAR JENIS U−7Mo/Al

    Directory of Open Access Journals (Sweden)

    Maman Kartaman Ajiriyanto

    2016-06-01

    Full Text Available ABSTRAK PENGARUH TEMPERATUR DAN IRADIASI TERHADAP INTERDIFUSI PARTIKEL BAHAN BAKAR JENIS U−7Mo/Al. Paduan U−7Mo/Al memiliki potensi besar sebagai bahan bakar reaktor riset, tetapi bahan bakar ini memiliki beberapa kekurangan antara lain dapat membentuk interaction layer pada antarmuka pada saat proses fabrikasi maupun iradiasi di reaktor melalui mekaniame difusi. Penelitian ini dilakukan untuk mengetahui terjadinya interaction layer yang disebabkan oleh interdifusi atau diffusion couple paduan U−7Mo dengan pelat AlMg2 yang dipanaskan pada temperatur 500 °C dan 550 °C selama 24 jam dalam tungku arc furnace dan tungku DTA pada temperatur 30 °C hingga 1400 °C. Hasil pengamatan mikrostruktur menggunakan Scanning Electron Microscope (SEM pada sampel diffusion couple hasil pemanasan pada temperatur 500 °C belum terlihat adanya interaction layeratau pembentukan fasa baru antara partikel U−Mo dan matriks Al. Sementara itu, pemanasan pada temperatur 550 °C telah terjadi interdifusi paduan U−7Mo dengan pelat AlMg2 menghasilkan senyawa (U,MoAlx pada antarmuka atau interface. Hal ini didukung oleh hasil analisis DTA menunjukkan bahwa paduan U−7Mo/Al pada 500 °C mempunyai kompatibilitas panas yang baik, tetapi diatas temperatur 550 °C telah terjadi perubahan fasa a + d menjadi a + g. Pemanasan hingga 679,14 °C terjadi fasa metastabil U(Al,Mox dan selanjutnya mengalami proses interdifusi dengan leburan uranium membentuk interaction layer berupa aglomerat senyawa UAlx (UAl4, UAl3 danUAl2. Aglomerat yang terbentuk dari proses pemanasan secara diffusion couple maupun dalam tungku DTA dibandingkan dengan aglomerat yang terbentuk akibat proses iradiasi. Bahan bakar paduan U−7Mo/Al yang diradiasi dengan burn up 58% mengalami interdifusi antara U−7Mo dengan matriks Al menghasilkan fasa metastabil U(Al,Mox yang berubah menjadi layer (U,MoAl7, presipitat UMo2Al20, (UMoAl3−Al dan membentuk boundary atau aglomerat UAlx (UAl4, UAl3 danUAl2

  16. Reactor Radiation Loops as Large Gamma Sources; Boucles d'irradiation des reacteurs nucleaires utilisees comme sources gamma intenses; Radiatsionnye kontury yadernykh reaktorov kak moshchnye gamma-istochniki; Empleo de circuitos de irradiacion de los reactores como fuentes gamma de gran intensidad

    Energy Technology Data Exchange (ETDEWEB)

    Ryabukhina, Yu. S.

    1963-11-15

    vybrany kontury na zhidkikh pri komnatnoj temperature splavakh indiya. Bylo izucheno povedenie dvukh ehvtekticheskikh splavov indiya po otnosheniyu k nekotorym konstruktivnym materialam, i v nachale 1960 g. byl zapushchen pervyj stendovyj indij-gallievyj kontur. V rezul'tate dal'nejshikh rabot byli zapushcheny model'nyj indij-galdievyj kontur s aktivnost'yu v obluchatele do {approx} 100 g. eh k v radiya pri reaktore IRT AN Gruzinskoj SSR i stendovyj indij-gallij-olovyannyj kontur v kanale reaktora IRT IAEH AN SSSR. Nakonets, v 1962 g. byl zapushen rabochij indij-gallij-olovyannyj kontur pri reaktore IRT AN Latvijskoj SSR dlya provedeniya polupromyshlennykh radiatsionnykh protsessov. Maksimal'naya aktivnost' obluchatelya kontura 30 000 g. eh k v radiya. Doklad sostoit iz 4 razdelov: 1. Raschety radiatsionnykh konturov. Obobshchaetsya prodelannaya rabota po metodam rascheta radiatsionnykh konturov. 2. Model' radiatsionnogo indij-gallievogo kontura reaktora IRT-2000 v Tbilisi. Opisyvaetsya dejstvuyushchij kontur. 3. Indij-gallij-olovyannyj radiatsionnyj kontur yadernogo reaktora IRT AN Latvijskoj SSR. Opisyvaetsya dejstvuyushchij kontur. 4. Perspektivy dal'nejshego razvitiya radiatsionnykh konturov. Opisyvayutsya ehksperimenty, skhemy i privodyatsya raschety, na osnovanii kotorykh predstavlyaetsya vozmozhnym sozdanie konturov na tverdom margantse i mobil'nykh konturov na zhidkikh splavakh indiya. (author)

  17. Assessment of End-Plug Welding of Fuel Elements; Evaluation des Soudures Terminales des Elements Combustibles; Otsenka kachestva privarki kontsevoj probki toplivnykh ehlementov; Inspeccion de la Soldadura del Tapon Terminal de los Elementos Combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Y.; Aoki, T. [Tokai Refinery, Atomic Fuel Corporation (Japan)

    1965-10-15

    investigacion Numero-Sign 3 (JRR-3). Ese reactor de 10 MW es moderado y refrigerado por agua pesada, y tiene elementos combustibles de uranio metalico revestidos de aluminio. Como entre el revestimiento y el alma hay solamente una union mecanica, puede producirse una tension en el tapon terminal como resultado del crecimiento del alma de uranio debido a la irradiacion. El ciclo termico produce tensiones analogas en las soldaduras. Como resultado de la diferencia de microestructura, las proximidades de estas que dan especialmente expuestas a la corrosion producida por el agua caliente. Mientras el reactor esta en servicio, es imprescindible asegurar su estanqueidad. Se han utilizado probetas especiales para estudiar la resistencia a la traccion, la fluencia a alta temperatura, los efectos del ciclo termico y la corrosion. Antes de hacer esos ensayos, y periodicamente durante su realizacion, se sometieron a examen no destructivo muchas clases de soldaduras y se verifico si habia escapes. La evaluacion de los resultados obtenidos puede servir para establecer normas de inspeccion, por ejemplo, mediante radiografia y examen visual de la soldadura del tapon. En la memoria se describen algunos otros resultados de ensayos efectuados con elementos combustibles revestidos de Magnox y Zircaloy. (author) [Russian] Ochen' vazhno ustanovit' sootnoshenie mezhdu rezul'tatami ispytanij i ispol'zovaniem ih v reaktore, a takzhe razrabotat' sami metody ispytanija bez razrushenija ispytyvaemogo ob{sup e}kta. Odnako sdelat' jeto dovol'no trudno, tak kak jeto svjazano s bol'shimi rashodami i bol'shoj radioaktivnost'ju. Bylo proizvedeno neskol'ko vidov ocenok vo vnereaktornom sostojanii s imitaciej vnutrireaktornyh uslovij. Opisyvajutsja nekotorye detali jetih ocenok v otnoshenii toplivnyh jelementov issledo- vatel'skogo reaktora JKK-3. V je t om reaktore ustanovlennoj moshhnost'ju 10 mgvt s tjazhe- loj vodoj v kachestve zamedlitelja i teplonositelja ispol'zujutsja toplivnye jelementy iz

  18. Improved Techniques for Low-Flux Measurement of Prompt Neutron Lifetime, Conversion Ratio and Fast Spectra; Methodes Perfectionnees de Mesure de la Duree de Vie des Neutrons Instantanes, du Rapport de Conversion et des Spectres de Neutrons Rapides, dans un Reacteur a Bas Flux; Usovershenstvovannye metody izmereniya vremeni zhizni mgnovennykh nejtronov, koehffitsienta konversii i spektra bystrykh nejtronov pri slabykh potokakh nejtronov; Tecnicas Perfeccionadas para la Determinacion del Periodo de los Neutrones Inmediatos, la Razon de Conversion y los Espectros de Neutrones Rapidos, con Flujos Reducidos

    Energy Technology Data Exchange (ETDEWEB)

    Armani, R. J.; Bennett, E. F.; Brenner, M. W.; Bretscher, M. M.; Cohn, C. E.; Huber, R. J.; Kaufmann, S. G.; Redman, W. C. [Argonne National Laboratory, Argonne, IL (United States)

    1964-02-15

    'sov schet- chika kak funkcii vremeni zapazdyvanija i izmerenie otnositel'nogo otklonenija integralov vo vremeni potoka nejtronov kak funkcii integriruemogo vremeni, byli uluchsheny. Izu- cheny oblasti naibolee uspeshnogo primenenija razlichnyh metodov. Bylo udeleno takzhe vni- manie interpretacii rezul'tatov jetih izmenenij, i bylo pokazano, chto interpretacija na osnove prostoj kineticheskoj modeli primenima dlja shirokogo kruga konkretnyh sluchaev. Opisyvajutsja neskol'ko usovershenstvovanij nashego pervonachal'nogo aktivacionnogo metoda opredelenija otnoshenija obrazovanija k razrusheniju deljashhegosja veshhestva pri slabyh potokah nejtronov. Sjuda vhodit primenenie ochen' chuvstvitel'nyh radiohimicheskih metodov dlja proverki poluchennyh rezul'tatov; vnesenie popravok s cel'ju ucheta pomeh; primenenie fol'g razlichnoj stepeni obogashhenija dlja dobavlenija aktivnosti deljashhegosja materiala k aktivnosti produktov delenija; primenenie fol'g. obluchennyh nejtronami razlichnyh jenergij, dlja scheta delenij pri opredelenii zahvata; i ispol'zovanie scheta sovpadenij dlja obnaruzhenija raspada Np{sup 239}. Dlja poluchenija znachenija kojefficienta konversii po dannym aktivacii ne- obhodimo znat' otnoshenie kolichestva zahvatov k kolichestvu aktov delenija v toplive ({alpha}). V nastojashhee vremja ne sushhestvuet tochnogo jeksperimental'nogo metoda dlja izmerenija jetoj velichiny v reaktore so slabym potokom nejtronov; issledujutsja neskol'ko metodov, koto- rye, vozmozhno, pozvoljat opredelit' jetu velichinu. Opisyvajutsja dva ili tri perspektivnyh metoda. Usilija po sozdaniju nebol'shogo spektrometra bystryh nejtronov, imejushhego horoshuju stabil'nost', razreshajushhuju sposobnost' i chuvstvitel'nost' dlja izmerenija v reaktore spekt- ra nejtronov v oblasti jenegrii v desjatki i sotni kilovol't, byli sosredotocheny na prime- nenii analiza formy impul'sov dlja iskljuchenija vyzyvaemyh gamma-luchami javlenij v vodo- rodnyh proporcional'nyh schetchikah otdachi i na

  19. Fabrication and Testing of Prototype APM-Clad UO{sub 2} Fuel Elements; Fabrication et essai de prototypes de cartouches de combustible en bioxyde d'uranium gaine d'aluminium (APM); Izgotovlenie i ispytanie prototipa toplivnykh ehlementov na osnove UO{sub 2} s obolochkoj iz alyuminiya metodom poroshkovoj metallurgii; Elaboracion y ensayo de elementos combustibles prototipo de UO{sub 2} con revestimiento de aluminio sinterizado

    Energy Technology Data Exchange (ETDEWEB)

    Ballif, III, J. L.; Friske, W. H.; Gordon, R. B. [Atomics International, Canoga Park, California (United States)

    1963-11-15

    los elementos combustibles se extrajeron del reactor a raiz de problemas suscitados por el instrumental utilizado para los ensayos. Los tres elementos restantes continuaban en el reactor. Todos los resultados de los experimentos realizados hasta la fecha parecen indicar que el sistema de combustible PMA-UO{sub 2} permitira alcanzar los objetivos fijados para el programa relativo al reactor POPR. (author) [Russian] V podderzhku Programmy prototipov organicheskikh ehnergoreaktorov (POPR) moshchnost'yu 50 mgvt (ehl.) byla provedena obshirnaya rabota po razrabotke poroshkovoj metallurgiej alyuminievykh materialov (APM) v kachestve obolochek dlya topliva iz UO{sub 2} . Kak chast' ehtoj raboty byli issledovany ehvteticheskoe soedinenie, svarke v styk oplavleniem i kholodnaya svarka. Kolduehlla pod davleniem kak metody dlya soedineniya stykov pri sborke toplivnykh ehlementov. Vibratsionnoe uplotnenie izuchalos' kak sredstvo zapolneniya trubok APM dvuokis'yu urana. Opyty vne reaktora provodilis' dlya.ogo, chtoby poluchit' informatsiyu o svomestimosti APM - UO{sub 2} . Ehta rabota pokazala, chto pri sushchestvuyushchikh usloviyakh ehvteticheskoe soedinenie yavlyaetsya naibolee podkhodyashchim sposobom dlya soedineniya stykov; vibratsionnoe uplotnenie davalo plotnost' topliva v predelakh ot 80 do 88% teoreticheskoj plotnosti, i ne nablyudalos' vzaimodejstvie APM -UO{sub 2} v diapazone ehkspluatatsionnykh temperatur POPR (temperatura poverkhnosti razdela toplivo - obolochka maksimum 850{sup o}F). V rezul'tate ehtoj raboty bylo izgotovleno 5 prototipov UO{sub 2}-APM toplivnykh ehlementov s tsel'yu ispytaniya na opytnom reaktore s organicheskim zamedlitelem. Kazhdyj ehlement sostoyal iz 24 ili 25 toplivnykh sterzhnej v obolochke iz APM i sgruppirovannykh v sistemu 5 x 5 v stal'noj korobke, pokrytoj nikelem, ili toplivnoj korobke ieh APM. Dlya togo, chtoby uvelichit' poverkhnost' ehlementa i tem samym znachitel'no uluchshit' teploperedachu, izgotovlennaya ehkstruziej obolochka ieh

  20. Operating Experience in Nuclear Power Plants with Boiling-Water Reactors; Experience acquise dans l'exploitation des reacteurs a eau bouillante; Opyt ehkspluatatsii kipyashchago reaktora; Experiencia adquirida con la explotacion de reactores de agua hirviente

    Energy Technology Data Exchange (ETDEWEB)

    Ascherl, R. J. [General Electric Company, San Jose, CA (United States)

    1963-10-15

    , ''Rejnish vestfalishes ehlektritsitetsverk und bajernverk'', Kal'-na-Majne, Zapadnaya Germaniya. Rabochaya kharakteristika kipyashchego reaktora atomnoj ehlektrostantsii pri obychnom rezhime raboty v kommunal'noj ehnergosisteme ochen' khoroshaya. Koehffitsient ispol'zovaniya i moshchnosti reaktora i ehlektrostantsii daet tverdoe osnovanie polagat', chto ehlektrostantsii s kipyashchimi reaktorami yavlyayutsya nadezhnymi s tochki zreniya ikh rabochej kharakteristiki. V techenie 1963 goda budut vvedeny v stroj chetyre dopolnitel'nye ehlektrostantsii s kipyashchimi reaktorami: atomnaya ehlektrostantsiya v Big Rok Pojnt, ''Kons'yumers pauehr kompani'', Sharl'vua, Michigan, atomnaya ehnergeticheskaya ustanovka v KHamboldt Bej, ''Pasifik gaz ehnd ehlektrik kompani'', Yurika, Kaliforniya, atomnaya ehlektrostantsiya v Garig'yano, Natsional'noe obshchestvo po atomnoj ehnergii, Skauri, Italiya,i Yaponskij demonstratsionnyj ehnergeticheskij reaktor. Yaponskij nauchno-issledovatel'skij institut po atomnoj ehnergii, Tokai-Mura, Yaponiya. Pusk i pervonachal'naya ehkspluatatsiya ehtikh ehlektrostantsij podtverzhdayut predpolozhenie o nadezhnosti ikh raboty, chto uzhe prodemonstrirovano atomnymi ehlektrostantsiyami v Drezdene, Kale i Vallesitose. Rabochaya kharakteristika atomnykh ehlektrostantsij v Drezdene, Kale i Vallesitose yavlyaetsya naglyadnym dokazatel'stvom stabil'nosti i bezopasnosti kipyashikh reaktorov. Krome togo, urovni radiatsii na samoj ehlektrostantsii i v okruzhayushchej srede znachitel'no nizhe predelov, ustanovlennykh litsenziyami na ehkspluatatsiyu. Podtverdilis' prostota i legkost' ehkspluatatsii kipyashchikh reaktorov. Kharakteristika kontrolya za nagruzkoj u kipyashchego reaktora s dvojnym tsiklom Drezdenskoj ehlektrostantsii okazalas' ochen' khoroshej. Krupnye i nebol'shie raboty po ukhodu i remontu mogut osushchestvlyat'sya obychnymi remontnymi gruppami bez vrednykh posledstvij ili bez limita vremeni, svyazannymi s soobrazheniyami radioaktivnogo oblucheniya. V

  1. Operating Experience with Indian Point Nuclear Electric Generating Station; Experience d'exploitation de la centrale nucleaire d'Indian point; Opyt ehkspluatatsii Indian-pojntskoj yadernoj ehlektrostantsii; Experiencia adquirida con la explotacion de la central nucleoelectrica de Indian point

    Energy Technology Data Exchange (ETDEWEB)

    Beattie, W. C.; Freyberg, R. H. [Consolidated Edison Company of New York, Inc., New York, NY (United States)

    1963-10-15

    'yu 275 tys. kvt. Toplivom v reaktore sluzhit smes' polnost'yu obogashchennoj okisi U{sup 235} i okisi Th{sup 232}. Stantsiya raspolozhena na reke Gudzon, primerno v 24 milyakh severnee N'yu-Jorka. V svyazi s takoj blizost'yu ot N'yu-Jorka v proekte stantsii byli predusmotreny dopolnitel'nye predokhranitel'nye ustrojstva dlya predotvrashcheniya otkloneniya reaktivnosti i radiatsionnykh ehffektov takogo otkloneniya. Stroitel'stvo zakoncheno v mae 1962 goda. Zagruzka topliva osushchestvlena v iyune, a 2 avgusta 1962 goda reaktor vpervye dostig kritichnosti. V techenie avgusta privodilis' ispytaniya ka maloj moshchnosti do 5 mgvt pri temperature okrukhaitsej sredy i povyshennoj temperature. 16 sentyabrya 1962 goda turbogenerator byl vpervye fazirovan s sistemoj Konsolidejtid Ehdison. Ispytanie reaktora na moshchnostyakh do 50% provodilos' do noyabrya i preryvalos' chastymi avtomaticheskimiostanovkami, bol'shaya chast' kotorykh byla vyzvana nepoladkami v neyadernoj chasti stantsii. Nepoladki v sisteme upravleniya privodov reguliruyushchikh sterzhnej byli samymi ser'eznymi pomekhami so storony yadernoj chasti stantsii dlya raboty s avtomaticheskim vvedeniem reguliruyushchikh sterzhnej i prichinoj zaderzhek s puskom reaktora posle avtomaticheskikh ostanovok. 14 noyabrya 1962 goda stantsiya byla ostanovlena dlya planovoj zameny truboprovodov v neyadernoj chasti stantsii i modifikatsij i dopolneniya sistemy provodov reguliruyushchikh sterzhnej. Poslednee vklyuchalo ustanovku sistemy ochistki sukhogo azota dlya kozhukhov privodov reguliruyushchikh sterzhnej, prednaznachennykh dlya svedeniya k minimumu ehffektov vody, prosachivayushchejsya cherez uplotneniya v kozhukhi reguliruyushchikh sterzhnej. Okazalos', chto ehto yavlyaetsya osnovnoj prichinoj nepravil'nykh pokazanij sistemy upravleniya reaktorom. Blok byl vnov' pushchen 1 yanvarya 1963 goda. Ispytanie reaktora na moshchnostyakh do 100% v usloviyakh stabil'noj nagruzki bylo zakoncheno 27 yanvarya 1963 goda. Rezul

  2. Handling and Separation of Short-Lived Radioisotopes from Research Reactors; Manipulation et Separation des Radioisotopes a Courte Periode Produits dans des Reacteurs de Recherche; ПОЛУЧЕНИЕ И ОТДЕЛЕНИЕ КОРОТКОЖИВУЩИХ ИЗОТОПОВ В ИССЛЕДОВАТЕЛЬСКИХ РЕАКТОРАХ; Manipulacion y Separacion de Radioisotopos de Periodo Corto Obtenidos en Reactores de Investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Meinke, W. W. [University of Michigan, Ann Arbor, MI (United States)

    1963-03-15

    iz-za udalennosti ot snabzhavshego reaktora chasto ogranichivalos' temi, kotorye imeet period poluraspada bol'she odnogo dnja. Jeto tormozilo issledovanija v oblasti proizvodstva i primenenija radioizotopov voobshhe i v to zhe vremja meshalo potrebitele tochno ocenit' vozmozhnosti mnogih korotkozhivushhih izotopov. V svjazi s nalichiem vo vsem mire issledovatel'skih reaktorov jeta zavisimost' ot mesta proizvodstva izotopov mozhet predstavljat' soboj izvestnuju trudnost', no v to zhe vremja ona daet vozmozhnost' provodit' rjad novyh jeksperimentov s korotkozhivushhimi izotopami. Dlja jetogo trebuetsja vplotnuju podojti k voprosam proizvodstva mechenyh atomov. Pochti v techenie pjati let programma raboty ka issledovatel'skom reaktore Michiganskogo universiteta vkljuchala obrabotku, pererabotku i izmerenie korotkozhivushhih izotopov. Obychno pol'zovalis' izotopami, prodolzhitel'nost' poluraspada kotoryh ravnjalas' chasam i dazhe minutam. Hotja v dannom sluchae glavnye usilija byli napravleny na issledovanija v oblasti aktivacionnogo analiza, ispol'zovavsheesja oborudovanie i metodika mogut byt' primeneny i dlja drugih issledovanij. Dlja poluchenija korotkozhivushhih izotopov ne objazatel'no imet' sovershennoe oborudovanie, bol'shie zapasy izotopov ili podderzhivat' trehsmennuju rabotu reaktora. Pri uverennom obrashhenii s prostymi manipuljatorami mozhno poluchit' luchshie rezul'taty, chem distancionnym upravleniem. Osnovoj sistemy, opisyvaemoj v doklade, javljaetsja prisposoblenie v vide pnevmaticheskoj truby, kotoraja dostavljaet obrazcy v vytjazhnoj kolpak laboratorii, primykajushhej k reaktoru, v techenie treh sekund posle obluchenija. Rastvorenie i bystroe otdelenie radiohimicheskim sposobom mozhet proizvodit'sja bez dal'nejshego peremeshhenija obrazca. Metody otdelenija, legko prisposablivaemye dlja kratkovremennoj shkaly (neskol'ko minut), vkljuchajut ne tol'ko selektivnuju jekstrakciju, anionnyj obmen i osazhdenie, no i takie novye metodiki, kak izotopnyj obmen i

  3. APLIKASI THERMAL PRE-TREATMENT LIMBAH TANAMAN JAGUNG (Zea mays SEBAGAI CO·SUBSTRAT PADA PROSES ANAEROBIK DIGESTI UNTUK PRODUKSI BIOGAS

    Directory of Open Access Journals (Sweden)

    Darwin Darwin

    2016-04-01

    mikroorganisme anaerobik untuk mengkonversi polimer yang berupa selulosa dan hemiselulosa menjadi biogas. Tujuan dari penelitian ini adalah untuk melakukan kajian mengenai penerapan thermal pre-treatment pada limbah tanaman jagung terhadap proses anaerobik digesi yang meliputi efisiensi proses digesi dan produksi biogas yang dihasilkan. Penelitian ini dilakukan dengan menggunakan reaktor tipe batch yang suhunya dipertahankan pada kondisi mesophilic atau di atas rata-rata suhu kamar (33 ± 2 oC. Hasil penelitian diperoleh bahwa thermal pre-treatment yang diberikan pada limbah tanaman jagung mampu mempercepat proses produksi biogas pada 10 hari pertama sehingga dapat mengurangi lag-phase pada proses anaerobik digesi. Limbah tanaman jagung yang diberikan thermal pre-treatment mengalami perlambatan produksi biogas pada hari ke 26 dengan rata-rata total produksi 12.412,5 mL untuk limbah tanaman jagung yang diberikan thermal pre- treatment selama 15 menit, dan 12.310 mL untuk limbah tanaman jagung yang diberikan thermal pre-treatment selama 25 menit, sedangkan limbah tanaman jagung yang tidak diberikan pre-treatment menghasilkan produksi biogas sebesar 12.557 mL pada hari ke 26. Produksi biogas harian tertinggi terjadi pada substrat yang diberikan thermal pre-treatment 25 menit, dengan produksi biogas tertinggi pada hari ke 9 dengan rata-rata produksi sebesar 915 mL. Substrat yang diberikan thermal pre-treatment 15 menit juga memproduksi biogas jauh lebih tinggi (772,5 mL pada hari ke 9 jika dibandingkan dengan substrat tanpa diberikan pre-treatment yang hanya memproduksi biogas sebesar 405 mL. Data hasil penelitian menunjukkan bahwa limbah tanaman jagung yang diberikan thermal pre-treatment memperoleh biogas yield lebih tinggi dari pada yang tidak diberikan pre-treatment dimana 670,39 mL/g volatile solids untuk thermal pre- treatment 15 menit, 690,65 mL/g volatile solids untuk thermal pre-treatment 25 menit dan 456,37 mL/g volatile solids untuk limbah tanaman jagung yang tidak

  4. Developments for transactinide chemistry experiments behind the gas-filled separator TASCA

    Energy Technology Data Exchange (ETDEWEB)

    Even, Julia

    2011-12-13

    auf dem Elektrodenmaterial ab. Wenn die Adsorption auf dem Elektrodenmaterial gegenueber der Adsorption auf einer Oberflaeche aus dem abzuscheidenen Element bevorzugt ist, so verschiebt sich das Potential zu hoeheren Werten und man spricht von Unterpotentialabscheidung. Moeglichkeiten automatisierter Elektrochemieexperimente hinter dem gas-gefuellten Separator TASCA wurden untersucht, um spaetere Studien mit Transaktiniden durchfuehren zu koennen. Der zweite Teil der Arbeit befasst sich mit der In-situ-Synthese fluechtiger Carbonylkomplexe mit Kernreaktionsprodukten. Spaltprodukte des Uran-235 und Californium-249 wurden am TRIGA Mainz Reaktor erzeugt und in kohlenstoffmonoxidhaltiger Atmosphaere thermalisiert. Die gebildeten fluechtigen Komplexe der Uebergangsmetalle konnten im Gasstrom transportiert werden. Desweiteren wurden kurzlebige Isotope der Elemente Wolfram, Rhenium, Osmium und Iridium am Linearbeschleuniger UNILAC am GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt, in Kernfusionreaktionen erzeugt und im gasgefuellten Separator TASCA vom Ionenstrahl und den Transferprodukten getrennt. Die Kernfusionprodukte wurden in TASCAs Fokalebene in einer Rueckstosskammer in einer Kohlenstoffmonoxid-Helium Gasmischung thermalisiert. Die so erzeugten Carbonyl-Komplexe kurzlebiger Isotope wurden mittels kernsprektroskopischer Methoden identifiziert und zum Teil gaschromatographisch untersucht. Anhand des Vergleichs mit Monte Carlo Simulationen wurde die Adsorptionsenthalpien auf Siliziumdioxid- und Goldoberflaechen bestimmt. Die Monte Carlo Simulationen basieren auf bereits existierenden Programmen und wurden den Geometrien der Chromatographiesaeulen entsprechend modifiziert. Alle ermittelten Adorptionsenthalpien - auf Silziumoxid sowie auf Gold - sind typisch fuer Physisorption. In einigen Faellen wurde auch die thermische Stabilitaet der Carbonylkomplexe untersucht. Hierbei zeigte sich, dass ab Temperaturen von ueber 200 C die Komplexe zerstoert werden. Es

  5. Evaluation of Dose: Comparative Effect of Fast Neutrons and other Types of Radiation on the Survival of E. Coli and S. Cerevisiae; Evaluation de la Dose Delivree et Actions Comparees des Neutrons Rapides et d'Autres Radiations sur la Survie de E. Coli et S. Cerevisiae; Otsenka dozy i sravnitel'noe vliyanie bystrykh nejtronov i drugikh vidov izlucheniya na vyzhivaemost' E. Coli i S. Cerevisiae; Evaluacion de la Dosis Suministrada y Comparacion de la Accion de los Neutrones Rapidos sobre la Supervivencia del E. Coli y del S. Cerevisiae con la de Otras Radiaciones

    Energy Technology Data Exchange (ETDEWEB)

    Arnaud, Y.; Bocquet, C. [Centre d' Etudes Nucleaires de Saclay (France)

    1964-05-15

    diverses radiations dependent de l'organisme vivant etudie, du critere biologique choisi, mais aussi de la dose d'irradiation avec laquelle on fait la comparaison. L'efficacite biologique relative apparait dans ces experiences comme etant egalement une fonction de la dose. L'action des neutrons rapides et des rayons X releve souvent de processus radiobiologique differents. Il est arbitraire de relier entre elles les doses de ces diverses radiations par des relations lineaires. (author) [Spanish] Junto al reactor EL3 se encuentra un convertidor de uranio que permite obtener neutrones rapidos. Con una bateria de camaras de fision se mide el flujo y la distribucion espectral de los neutrones rapidos. Estos micro- detectores estan colocados en diversos puntos del volumen que se ha de irradiar y permiten evaluar experimentalmente la dosis absorbida en los tejidos. Gracias a este dispositivo y a este sistema de dosimetria, los autores han podido comparar la accion de los neutrones rapidos con las de otras radiaciones ionizantes (rayos X, rayos y) sobre organismos unicelulares. En este trabajo, estudian el porcentaje de supervivencia y la frecuencia de una mutacion morfologica en el Saccharomyces cerevisiae. Tambien han trazado la curva de supervivencia del Escherichia coli, expuesto a rayos X y a neutrones. La observacion de los resultados indica que los efectos biologicos relativos de estas diversas radiaciones dependen del organismo vivo estudiado, del criterio biologico elegido y, ademas, de la dosis de irradiacion con que se realiza la comparacion. Tambien se deduce de estos experimentos que la eficacia biologica relativa es funcion de la dosis. A menudo, la accion de los neutrones rapidos y de los rayos X depende de diferentes procesos radiobiologicos. Seria arbitrario establecer relaciones lineales entre las dosis de dichas radiaciones. (author) [Russian] Na reaktore EL-3 imeetsja uranovyj preobrazovatel, pozvoljajushhij poluchat' bystrye nejtrony. Batareja kamer delenija

  6. PENGARUH SERBUK U-Mo HASIL PROSES MEKANIK DAN HYDRIDE – DEHYDRIDE – GRINDING MILL TERHADAP KUALITAS PELAT ELEMEN BAKAR U-Mo/Al

    Directory of Open Access Journals (Sweden)

    Supardjo Supardjo

    2015-07-01

    Full Text Available PENGARUH SERBUK U-Mo HASIL PROSES MEKANIK DAN HYDRIDE – DEHYDRIDE – GRINDING MILL TERHADAP KUALITAS PELAT ELEMEN BAKAR U-Mo/Al. Penelitian bahan bakar U-7Mo/Al tipe pelat dilakukan dalam rangka pengembangan bahan bakar U3Si2/Al untuk mendapatkan bahan bakar baru yang memiliki densitas uranium lebih tinggi, stabil selama digunakan sebagai bahan bakar di dalam reaktor dan mudah dilakukan proses olah ulangnya. Lingkup penelitian meliputi pembuatan: paduan U-7Mo dengan teknik peleburan, pembuatan serbuk U-7Mo dengan dikikir dan hydride - dehydride - grinding mill, IEB U-7Mo/Al dengan teknik kompaksi pada tekanan 20 bar, dan PEB U-7Mo/Al dengan teknik pengerolan panas pada temperatur 425oC. Paduan U-7Mo hasil proses peleburan cukup homogen, berat jenis 16,34 g/cm3 dan bersifat ulet, kemudian dibuat menjadi serbuk dengan cara dikikir dan hydride - dehydride - grinding mill. Serbuk U-7Mo hasil proses kikir berbentuk pipih, kontaminan Fe cukup tinggi, sedangkan serbuk hasil proses hydride - dehydride - grinding mill, cenderung equiaxial dengan kontaminan yang rendah. Kedua jenis serbuk U-7Mo tersebut digunakan sebagai bahan baku pembuatan IEB U-7Mo/Al dan PEB U-7Mo/Al dengan densitas uranium 7 gU/cm3 dan diperoleh produk dengan kualitas yang hampir sama. Hasil uji IEB U-7Mo/Al berukuran 25 x 15 x 3,15±0,05 mm, tidak terdapat cacat/retak, distribusi U-7Mo di dalam matriks cukup homogen dan tidak terdapat pengelompokan/aglomerasi U-7Mo yang berdimensi >1 mm. PEB U-7Mo/Al hasil pengerolan dengan tebal akhir 1,45 mm, memiliki ketebalan meat rerata 0,60 mm dan tebal kelongsong 0,4 mm dan terdapat 1 titik pengukuran kelongsong dengan ketebalan 0,15 mm. Dengan membandingkan penggunaan kedua jenis serbuk U-7Mo tersebut, IEB U-7Mo/Al dan PEB U-7Mo/Al yang dihasilkan memiliki kualitas hampir sama. Namun demikian penggunaan serbuk U- 7Mo hasil proses hydride - dehydride - grinding mill lebih baik karena proses pengerjaannya lebih cepat dan impuritas dalam

  7. Imobilisasi TiO2 ke dalam Resin Penukar Kation dan Aplikasinya sebagai Fotokatalis dalam Proses Fotoreduksi Ion Hg2+

    Directory of Open Access Journals (Sweden)

    Rosyid Ridho

    2017-03-01

    Full Text Available Abstrak Dalam rangka mengembangkan bahan fotokatalitis TiO2 pada penelitian ini telah dilakukan preparasi fotokatalis TiO2-Resin yang disertai dengan karakterisasi dan uji aktivitas untuk proses fotoreduksi ion Hg(II. Preparasi imobilisasi ini dilakukan dengan metode pertukaran ion yang di ikuti dengan kalsinasi pada suhu tertentu. Pada preparasi telah dipelajari pengaruh konsentrasi Titanium Isopropoksida sebagai sumber ion Ti(IV terhadap TiO2-Resin yang dikarakterisasi dengan menggunakan Difraksi Sinar X (XRD dan Thermografimetri (TGA. Pada proses fotoreduksi ion Hg(II dipelajari pengaruh massa fotokatalis, kadar TiO2 yang terimobilisasi ke dalam resin, konsentrasi Ion Hg(II, dan pengaruh pH. Proses fotoreduksi dilakukan dalam suatu reaktor tertutup yang dilengkapi dengan lampu UV, yaitu dengan cara menyinari campuran yang terdiri dari larutan ion Hg(II dan serbuk fotokatalis TiO2-Resin, disertai dengan pengadukan selama waktu tertentu. Hasil fotoreduksi dihitung berdasarkan selisih antara konsentrasi ion Hg(II awal dengan ion Hg(II yang tak tereduksi. Penentuan konsentrasi ion Hg(II yang tak tereduksi dilakukan dengan menggunakan Spektrofotometer Serapan Atom (SSA teknik pembangkitan uap dingin atau Cold Vapor Atomic Absorption Spectrophotometry(CV-AAS. Hasil preparasi menunjukkan semakin tinggi konsentrasi Titanium Isopropoksida yang ditambahkan pada resin semakin tinggi juga kadar TiO2 yang terbentuk pada TiO2-Resin. Hasil uji fotokatalis menunjukkan bahwa penggunaan fotokatalis TiO¬2-Resin dapat meningkatkan hasil fotoreduksi ion Hg(II yang peningkatannya lebih tinggi dibandingkan TiO2 serbuk. Penambahan fotokatalis dengan massa yang semakin besar menambah efektivitas fotoreduksi terhadap ion Hg(II yang semakin besar, namun jika ditambahkan massa fotokatalis yang lebih tinggi lagi akan menurunkan efektivitas fotoreduksi terhadap ion Hg(II. Kenaikan konsentrasi Hg(II menyebabkan efektivitas fotoreduksi semakin rendah. Pada pH 1-4 terjadi

  8. Use of Neutron Irradiations in the Brookhaven Mutations Programme; Irradiation Neutronique dans le Cadre du Programme de Mutations Radioinduites de Brookhaven; Primenenie nejtronnogo izlucheniya v brukkhejvenskoj programme po ispol'zovaniyu mutatsij; La Irradiacion Neutronica en el Marco del Programa de Mutaciones Radioinducidas de Brookhaven

    Energy Technology Data Exchange (ETDEWEB)

    Miksche, J. P.; Shapiro, S. [Biology Department, Brookhaven National Laboratory, Upton, NY (United States)

    1964-03-15

    colaboracion con expertos de Australia, Belgica, Costa Rica, Chile, Dinamarca, Ecuador, Formosa, Grecia, Guatemala, India, Irlanda, Italia, japon, Kenia, Mexico, Paises Bajos, Pakistan, Peru, Filipinas, Rumania, Sudafrica, Tailandia, Venezuela, Alemania Occidental y Yugoeslavia. Los autores presentaran una resefla de los proyectos precitados, deteniendose ante todo en el uso de los neutrones para inducir mutaciones. EBR, por ejemplo, la capacidad de restablecimiento y la manifestacion del efecto oxigeno, principalmente para dosis correspondientes a valores reducidos de la TLE. Si bien esta interpretacion debe considerarse provisional, la distribucion de la dosis en funcion de la TLE proporciona una base para la realizacion de nuevos experimentos sobre la relacion existente entre la EBR y la TLE. (author) [Russian] Programma sotrudnichestva po ispol'zovaniju radiomutacij byla razrabotana v Brukhejvenskoj nacional'noj laboratorii primerno desjat' let nazad, chtoby predostavit' vozmozhnost' rastenievodam i agronomam primenjat' metody obluchenija po programme uluchshenija sortov rastenija. V kachestve ustanovki dlja obluchenija po programme ispol'zovalis' teplovaja kolonna v Brukhejvenskom grafitovom jeksperimental'nom reaktore, rentgenovskij apparat (pikovoe naprjazhenie 250 kv) biologicheskogo otdela jadernoj tehniki, gamma-istochnik moshhnost'ju 12 kjuri v teplicah i istochnik Co{sup 60}, raspolozhennyj v pole ploshhad'ju 13 akrov. V jetoj programme sotrudnichestva na dolju Brukhejvenskoj laboratorii prihoditsja razrabotka oborudovanija metodov i teoreticheskih polozhenij, v to vremja kak rastitel'nyj material i semena predostavljajutsja jekspertami po sel'skomu hozjajstvu, o t vetstvennymi za vyrashhivanie obluchennogo materiala i otbor mutacij. Bo le e 150 uchenyh iz 45 shtatov i Pujerto-Riko uchastvujut v vypolnenii programmy. Nachato takzhe vypolnenie proektov s Avstraliej, Bel'giej, Chili, Ko sta'Rika, Daniej, Jekvadorom, Tajvanem, Greciej, G vatem a loj, Indiej

  9. The use of radioactive inserts in the study of metal deformation during tube-making processes; Emploi de fils metalliques radioactifs pour l'etude des deformations des metaux pendant la fabrication de tubes; Ispol'zovanie radioaktivnykh vtulok dlya izucheniya deformatsii metallov pri protsesse izgotovleniya trub; Empleo de insertos radiactivos en el estudio de la deformacion de los metales durante la fabricacion de tubos

    Energy Technology Data Exchange (ETDEWEB)

    Davison, W H.T. [Tube Investments Research Laboratories, Cambridge (United Kingdom)

    1962-01-15

    experimental, que es aplicable a los estudios sobre deformacion de los metales en general. En este metodo es preciso localizar el inserto radiactivo con un error no mayor de un milimetro, lo que exige un poder de resolucion superior al que requieren normalmente las tecnicas de exploracion medica. Se examinan los problemas correspondientes en terminos de energia gamma, diseno del colimador y discriminacion de energias. (author) [Russian] V nastoyashchej rabote rassmatrivaetsya vopros ob ispol'zovanii radio- aktivnykh vtulok pri izuchenii deformatsii stali v pirsing-protsesse s pomoshch'yu goryachego rotatsionnogo metoda. ZHeleznaya ili stal'naya provoloka aktiviruetsya v reaktore, pomeshchaetsya v truby, pro- sverlennye v zheleznykh ili stal'nykh zagotovkakh, kotorye zatem podvergayutsya obrabotke v normal'nykh proizvodstvennykh usloviyakh. Tshchatel'noe nablyudenie pokazyvaet, chto radiatsionnoe obluchenie i zagryazneniya ustanovki sovershenno neznachitel'ny. Posle okhlazhdeniya zagotovki podvergayutsya analizu v laboratorii putem primeneniya kollimirovannogo stsintillyatsionnogo schetchika, raspolozhennogo perpendikulyarno k poverkhnosti prosverlennoj zagotovki; krivye izodoz pokazyvayut formu deformiro- vannoj radioaktivnoj vtulki, raspolozhennoj pod pryamym uglom k poverkhnosti truby. Drugoj grafik poluchaetsya ot podobnogo roda razvertki uchastka, nakhodyashchegosya pod pryamym uglom k osi truby; pri pomoshchi dvukh poluchennykh grafikov poluchaetsya kartina v trekh izmereniyakh. V dokumente privodyatsya nekotorye rezul'taty, no osnovnoe vnimanie v nem udelyaetsya metodam provedeniya ehksperimentov, kotorye shiroko ispol'zuyutsya pri izuchenii defor- matsii metallov. Vtulki dolzhny byt' raspolozheny na rasstoyanii millimetra ili okolo ehtogo, dlya chego trebuetsya bolee vysokaya razreshayushchaya sposobnost' po sravneniyu s normal'nymi metodami issledovaniya, primenyaemymi v meditsine. Problemy dostizheniya ehtogo ras- smatrivayutsya v nastoyashchem doklade v zavisimosti

  10. The Problem of Storing Fission Products Arising from the Processing of Irradiated Uranium-Molybdenum Alloys; Probleme du Stockage des Produits de Fission en Provenance du Traitement des Alliages Uranium-Molybdene Irradies; 041f 0420 041e 0411 041b 0415 041c 0410 0425 0420 0410 041d 0415 041d 0418 042f 041e 0422 0425 041e 0414 041e 0412 041f 041e 0421 041b 0415 041f 0415 0420 0415 0420 0410 0411 041e 0422 041a 0418 041e 0411 041b 0423 0427 0415 041d 041d 041e 0413 041e 0421 041f 041b 0410 0412 0410 0423 0420 0410 041d - 041c 041e 041b 0418 0411 0414 0415 041d ; El Problema del Almacenamiento de los Productos de Fision Procedentes del Tratamiento de las Aleaciones Uranio-Molibdeno Irradiadas

    Energy Technology Data Exchange (ETDEWEB)

    Faugeras, P.; Kikindai, T. [Commissariat a l' Energie Atomique, Paris (France)

    1963-02-15

    el reactor, pero presentan serios inconvenientes en lo que respecta al almacenamiento de los productos de fision resultantes de su tratamiento. La insolubilidad del molibdeno impide concentrar por evaporacion las soluciones de los productos de fision. Por tal motivo, los autores han estudiado la posibilidad de solubilizar el molibdeno agregandole reactivos tales como el hierro o los iones fosforicos. De este modo, es posible obtener soluciones finales de 60 g/l Mo con hierro, 100 g/l Mo con PO{sub 4}H{sub 3}. Los volumenes que es preciso almacenar siguen siendo importantes (sobre todo con el hierro) y por eso se ha estudiado la calcinacion de los nitratos en un lecho fluidizado. La reaccion tiene lugar a los 400 Degree-Sign C aproximadamente. El comportamiento del rutenio y la friabilidad del solido calcinado (importante formacion de particulas menudas) ha inducido a los autores a reemplazar este procedimiento por el de preparacion de vidrios fosfAtados. (author) [Russian] Povedenie splavov uran-molibden v reaktore predstavljaet prakticheskij interes. Odnako voznikaet problema hranenija produktov delenija, kotorye poluchajutsja v rezul'tate ih pererabotki. V tom sluchae, kogda molibden nahoditsja v nerastvorennom sostojanii, ljuboe koncentrirovanie produktov delenija putem isparenija rastvora prakticheski nevozmozhno. Vot pochemu my orientirovalis' na izuchenie rastvorenija mo'libdena pribavleniem takih reagentov, kak zhelezo idi fosfat-iony. Mozhno poluchit' rastvor sostava: 60 g/l Mo s Re ili. 100 g/l Mo sPO{sub 4}H{sub 3} V jetom sluchae ob'edy, prednaznachennye dlja hranenija, uvelichivajutsja (osobenno pri dobavlenii Fe). Bylo izucheno prokalivanie nitratov v dvizhushhemsja sdoe. Reakcija prohodit pri temperature okolo 400 Degree-Sign C. Letuchest' rutenija i hrupkost' tverdogo prokalennogo veshhestva (vsledstvie obrazovanija znachitel'nogo kolichestva uglja) zastavila'nas otkazat'sja ot'jetogo processa i'perejti k izgotovleniju fosfatnyh stekol. (author)

  11. Critical experiments and nuclear calculations - LAMPRE-I; Experiences critiques et calculs nucleaires concernant le LAMPRE-I; Kriticheskie opyty i yadernye raschety - LAMPRE-I; Experimentos criticos u calculos nucleares relativos al LAMPRE-I

    Energy Technology Data Exchange (ETDEWEB)

    Battat, M E [Los Alamos Scientific Laboratory, University of California, Los Alamos, NM (United States)

    1962-03-15

    utilise la methode S{sub n} pour resoudre le probleme du transport neutronique. La comparaison entre les valeurs calculees et les valeurs mesurees des parametres, tels que le coefficient thermique, l'efficacite des barres de controle et la masse critique, presente aussi de l'interet pour evaluer le degre de confiance que l'on peut accorder aux calculs des bureaux d'etudes. (author) [Spanish] Como parte de un programa de ensayos de combustibles de plutonio para reactores reproductores de neutrones rapidos, se ha construido y puesto en marcha en el Los Alamos Scientific Laboratory un reactor experimental de 1 MW refrigerado por sodio, cuyo cuerpo contiene una aleacion fundida de plutonio y hierro (90 atomos por ciento de Pu y 10 atomos por ciento de Fe; punto de fusion: 410 deg. C). La reactividad se regula por medio de un reflector de acero inoxidable y de cuatro barras de control de niquel situadas fuera del nucleo. Se han llevado a cabo experimentos a temperaturas (isotermicas) de 80, 160 y 480 deg. C en el cuerpo, a fin de determinar la masa critica y la eficacia del reflector a cada una de esas temperaturas. Tambien se midio la eficacia de las barras de control, por registro de los periodos y del coeficiente termico de la reactividad. Aplicando el metodo S{sub n} de resolucion del problema del transporte neutronico, se efectuaron calculos para determinar los parametros nucleares basicos del reactor. La comparacion entre los valores calculados y los valores medidos de parametros tales como el coeficiente termico, la eficacia de las barras de control y la masa critica, presenta tambien interes en lo que se refiere a la evaluacion del grado de confianza que puede atribuirse a los calculos del proyectista. (author) [Russian] V kachestve chasti programmy po razvitiyu plutonievogo topliva dlya reaktorov-razmnozhitele j na bystrykh nejtronakh Los-Alamosskaya nauchnaya laboratoriya skonstruirovala i ehkspluatiruet ispytatel'nyj reaktor s natrievym okhlazhdeniem moshchnost'yu v

  12. Presentation of the results for deuterium retention and thermal release in a new type of low activation ferritic-martensitic steel EUROFER / Результаты исследования по удержанию дейтерия и термической десорбции в условиях низкой активации ферритно-мартенситной стали EUROFER / Rezultati zadržavanja i termalne desorpcije deuterijuma u EUROFER-u, novoj vrsti feritno-martenzitnog čelika niske aktivacije

    Directory of Open Access Journals (Sweden)

    Sanja Lj. Korica

    2016-04-01

    žđu, hromu i EOROFER-u, leguri koja se razmatra kao najnoviji materijal za buduće fuzione reaktore. Studija je pokazala sledeće rezultate: zadržavanje deuterijuma u hromu je mnogo veće nego u gvožđu (usled formiranja hidrida hroma, zadržavanje deuterijuma u EUROFER-u je za faktor 2 veće nego u gvožđu, primećena je specifična struktura u koncentracionom profilu gvožđa i EUROFER-a na dubini ~ 4 μm, veliki stepen difuznosti i zadržavanja deuterijuma govore o potencijalnoj upotrebi Au kao difuzione barijere u fuzionom reaktoru.

  13. The Effect of the Ammonium Group on the Different Annealing Processes; Role du Groupe Ammonium dans les Differents Processus de Recuit; 042d 0424 0424 0415 041a 0422 0410 041c 041c 041e 041d 0418 0415 0412 041e 0419 0413 0420 0423 041f 041f 042b 041d 0410 0420 0410 0417 041b 0418 0427 041d 042b 0415 041f 0420 041e 0426 0415 0421 0421 042b 041e 0422 0416 0418 0413 0410 ; Efecto del Grupo Amonio en Diferentes Procesos de Recocido

    Energy Technology Data Exchange (ETDEWEB)

    Getoff, N. [Institute Of Chemistry, Oesterreichische Studiengesellschaft fuer Atomenergie GmbH, Vienna (Austria)

    1965-04-15

    , fotohimicheskogo i ul'trazvukovogo otzhiga sul'fata ammonija i kalija, a takzhe dihromatov ammonija i kalija. Zatem predstavljajutsja jeksperimental'nye dannye, pokazyvajushhie vlijanie ammonievoj gruppy na teplovoj, radiacionnyj i ul'trazvukovoj otzhig mono-, di- i trifosfatov ammonija. Bylo ustanovleno, chto v uslovijah nejtronnogo obluchenija v reaktore ammonievaja gruppa soedinenij mozhet transformirovat'sja v nekotorye metastabnl'nye fragmenty otdachi, kotorye mogut privesti k obrazovaniju vosstanavlivajushhih veshhestv, naprimer gidrazina. Vyhod gidrazina uvelichivaetsja pri posledujushhej obrabotke obluchennyh nejtronami tverdyh obrazcov gamma-luchami, ul'trazvukom i t.d. Odnovremenno s rostom vyhoda gidrazina proishodit bol'shoe izmenenie v uderzhanii. (author)

  14. A modification of the method for determining current efficiency of aluminium electrolytic cells; Modification de la methode permettant de determiner le rendement des cuves dans la production d'aluminium par electrolyse; Izmenenie metoda opredeleniya ehffektivnosti toka v alyuminievykh ehlektroliticheskikh bakakh; Modificacion del metodo para determinar el rendimiento de las celdas utilizadas en la produccion de aluminio por electrolisis

    Energy Technology Data Exchange (ETDEWEB)

    Pradzynski, A [Institute of Basic Technical Problems, Polish Academy of Sciences. Warsaw (Poland); Orman, Z [Institute of Nonferrous Metals, Gliwice (Poland)

    1962-01-15

    alyuminievykh pechakh, opisannyj vpervye REMPELEM i dr., byl usovershenstvovan BOZUKI i dr. pri pomoshchi ispol'zovaniya radioaktivnogo izotopa zolota Au{sup 198}. Pri provedenii upomyanutogo issledovaniya byli izgotovleny obraztsy splavov alyuminiya s zolotom Au{sup 198}, i proby ehtikh obraztsov s vysokoj udel'noj aktivnost'yu byli izmereny pri pomoshchi schetchika Gejgera-Myullera so svintsovym poglotitelem, vstavlennym mezhdu schetchikom i proboj. Udel'naya aktivnost' obraztsa splava byla izmerena posle razbavleniya ego opredelennym kolichestvom chistogo alyuminiya. Takim obrazom proby razbavlennogo splava i proby, vzyatye iz ehlektroliticheskogo baka, imeli udel'nuyu aktivnost' togo zhe poryadka velichiny i mogli izmeryat'sya bez vsyakogo poglotitelya. Dlya togo, chtoby oblegchit' primenenie ehtogo metoda na alyuminievom zavode i vo izbezhanie vsyakikh ogranichenij i opasnosti, svyazannykh s obrashcheniem s otkrytymi radioaktivnymi istochnikami vne osobykh laboratorij, prednaznachennykh dlya issledovaniya radioaktivnykh izotopov, byl ispol'zovan radioaktivatsionnyj analiz. K obraztsu splava, a takzhe i v nakhodyashchuyusya v ehlektroliticheskom bake plavil'nuyu vannu bylo dobavleno neaktivirovannoe zoloto. Kontsentratsiya zolota kak v obraztsakh splava, tak i v plavil'noj smesi byla zatem izmerena posle oblucheniya prob v yadernom reaktore. (author)

  15. Post-Irradiation Behaviour of I{sup 131} in TeO{sub 2}; Comportement de {sup 131}I Dans TeO{sub 2} Apres Irradiation; 041f 041e 0412 0415 0414 0415 041d 0418 0415 0419 041e 0414 0410 -131 0412 TeO{sub 2} 0433 041f 041e 0421 041b 0415 041e 0411 041b 0423 0427 0415 041d 0418 042f ; Comportamiento del {sup 131}I en TeO{sub 2} Despues de su Irradiacion

    Energy Technology Data Exchange (ETDEWEB)

    Jacimovic, Lj.; Stevovic, J.; Veljkovic, S. R. [Boris Kidric Institute of Nuclear Sciences, Belgrade, Yugoslavia (Serbia)

    1965-04-15

    cinetica del recocido. (author) [Russian] Sistema joda-131 v TeO{sub 2} predstavljaet interes, potomu chto malo izvestno o termohimicheskih izmenenijah v jetoj misheni. Radioaktivnyj jod poluchalsja pri nejtronnom obluchenii TeO{sub 2} v reaktore. Obluchennyj TeO{sub 2} rastvorjalsja v razbavlennom rastvore NaOH. Analiz razlichnyh valentnyh form joda proizvodilsja ionoobmennym metodom. Izuchalis' termicheskaja i radiacionnaja ustojchivost' TeO{sub 2} putem ispol'zovanija spektrofotometricheskogo metoda dlja opredelenija tellura i otzhig joda-131 v TeO{sub 2} posle obluchenija v zavisimosti ot vremeni i temperatury nagreva. Glavnaja tendencija otzhiga sostojala v vosstanovlenii radioaktivnogo joda. Zavisimost' jetogo processa ot vremeni ukazyvaet na bystroe izmenenie pri vysokoj temperature. Pri bolee nizkih temperaturah krivye javljajutsja bolee slozhnymi. Otzhig verojatno javljaetsja slozhnym vvidu mnogoobrazija teplovyh reakcij, vkljuchajushhih promezhutochnye soedinenija joda. On mozhet vstupat' v reakciju s produktami processa otdachi jader tellura v sootvetstvujushhej gorjachej zone, beta-perehoda Te{sup 131} i s samoj dvuokis'ju tellura. Rassmatrivalas' kinetika jetih izmenenij i proizvodilas' ocenka processov vo vremja otzhiga. Izuchalos' takzhe vlijanie nejtronnogo potoka na kinetiku otzhiga. (author)

  16. The Scottish Research Reactor Centre and its Facilities for the Production and Exploitation of Short-Lived Radioisotopes; Le Réacteur de Recherche Ecossais et ses Installations pour la Production et l'Exploitation des Radioisotopes a Courte Periode; ШОТЛАНДСКИЙ ИССЛЕДОВАТЕЛЬСКИЙ РЕАКТОРНЫЙ ЦЕНТР И ЕГО ТЕХНИЧЕСКИЕ СРЕДСТВА ДЛЯ ПРОИЗВОДСТВА И ИСПОЛЬЗОВАНИЯ КОРОТКОЯИВУШХ РАДИОИЗОТОПОВ; El Centro del Reactor de Investigacion de Escocia y sus Instalaciones para la Produccion y Empleo de Radioisotopos de Periodo Corto

    Energy Technology Data Exchange (ETDEWEB)

    Ward, A. [The Royal College of Science and Technology, Glasgow, Scotland (United Kingdom)

    1963-03-15

    , cuartos oscuros, laboratorios para trabajar con actividades del orden del microcurie, criaderos de animales, laboratorios biologicos y quimicos, local de recuento de baja actividad de fondo, sala de conferencias y biblioteca. Se espera que las investigaciones abarquen muchos campos cientificos y tecnologicos. Gran parte de los trabajos se efectuaran con ayuda de radioisotopos de periodo corto. La memoria describe algunos de los proyectos mas caracteristicos. (author) [Russian] V nastojashhee vremja vedetsja stroitel'stvo Shotlandskogo issledovatel'skogo reaktornogo centra, kotoryj vstupit v stroj k vesne 1963 goda. Ustanovka predstavljaet soboj reaktor bakovogo tipa moshhnost'ju 100 kvt s vodjanym ohlazhdeniem i vodografitovym zamedlitelem pri ispol'zovanii obogashhennogo topliva (uran{sup 235}). Jeksperimental'noe tehnicheskoe oborudovanie vkljuchaet bol'shuju teplovuju kolonnu., bol'shoj vodjanoj bak s zashhitoj, a takzhe ustanovku dlja proizvodstva radioizotopov s pnevmaticheskim transporterom. V centre aktivnoj zony imejutsja tri skvoznye truby; imeetsja takzhe odna skvoznaja truba v teplovoj kolonne,a takzhe neskol'ko nebol'shih central'nyh vertikal'nyh stringerov i odin vertikal'nyj stringer secheniem v dkjmov, dohodjashhie do centra aktivnoj zony reaktora. Mnogo gorizontal'nyh stringerov prohodit cherez teplovuju kolonnu, prichem central'nyj stringer prohodit na rasstojanii odnogo dkjma ot toplivnogo baka. V dopolnenie k reaktornomu tehnicheskomu oborudovanie imeetsja mnozhestvo nebol'shih laboratorij. Jeti laboratorii vkljuchajut oborudovanie dlja poluchenija gorjachih istochnikov i raboty s nimi, a takzhe komnatu dlja pereodevanija, jelektricheskie i mehanicheskie masterskie, temnye komnaty, laboratorii dlja raboty s mikrokjurievymi kolichestvami, pomeshhenie dlja zhivotnyh, biologicheskuju i himicheskuju laboratorii, schetnuju komnatu s nizkim urovnem fona, lekcionnyj zal i biblioteku. Predpolagaetsja, chto issledovanija budut kasat'sja mnogih nauchnyh i

  17. Control Rods in high-Flux Swimming-Pool Reactors; Les Barres de Controle dans les Piles Piscines a Haut Flux; Reguliruyushchie sterzhni dlya reaktorov bassejnovogo tipa s vysokoj plotnost'yu nejtronnogo potoka; Las Barras de Control en los Reactores Tipo Piscina de Flujo Elevado

    Energy Technology Data Exchange (ETDEWEB)

    Ageroni, P.; Blum, P.; Denielou, G.; Denis, P.; Meunier, C. [Centre d' Etudes Nucleaires de Grenoble (France)

    1964-06-15

    i kombinacii jetih razlichnyh jelementov); b) piki potokov, sozdavaemyh v aktivnoj zone prisutstviem regulirujushhih sterzhnej, iz vlijanie na udel'nuju moshhnost', potoki bystryh nejtronov, kotorye mogut byt' polucheny takim obrazom, a takzhe sposoby usilenija jetih potokov; v) tehnologicheskie problemy, voznikajushhie v svjazi s izgotovleniem regulirujushhih sterzhnej; g) problemy ohlazhdenija, vibracii, deformacii i vremeni bystrogo zakrytija na dejstvujushhem reaktore. 3. V zakljuchenie daetsja kratkij obzor provodimogo v nastojashhee vremja izuchenija regulirujushhih sterzhnej dlja reaktorov bassejnovogo tipa s otkrytoj aktivnoj zonoj, rabotajushhih v diapazone moshhnostej 10 - 30 mgvt. (author)

  18. Advanced epithermal thorium reactor (AETR) physics; Physique d'un reacteur au thorium, a neutrons epithermiques, de type perfectionne (AETR); Fizika usovershenstvovannog o nadteplovogo torievogo reaktora; Fisica del reactor epitermico de tipo avanzado, alimentado con torio (AETR)

    Energy Technology Data Exchange (ETDEWEB)

    Campise, A. V. [Atomics International, Canoga Park, CA (United States)

    1962-03-15

    del {sup 233}Pa y de isotopos del uranio sobre el balance neutronico relativo y se evalua la probable razon de reproduccion y las caracteristicas de combustion teniendo en cuenta la imprecision en el conocimiento de las secciones eficaces nucleares. (author) [Russian] Rassmatrivayuts ya printsipy konstruirovani ya usovershenstvovannog o nadteplovogo torievogo reaktora s uchetom sushchestvuyushchej teorii yadernykh parametrov i potentsial'nogo poleznogo ispol'zovaniya nejtronov. Byl izuchen ehffekt rezonansnogo zakhvata toriya v sistemakhs grafitovym zamedlitelem dlya nejtronov s ehnergiyami ot 0,10 do 100 kehv. Ispol'zuyutsya formuly uzkogo rezonansa i shirokogo rezonansa v tselyakh polucheniya zavisimogo ot temperatury ehffektivnogo rezo- nansnogo integrala torievogo sterzhnya, kotoryj vyrazhaetsya v vide ehkvivalentnykh mnogogruppovykh sechenij. Neobkhodimost' v poluchenii yadernykh dannykh v oblasti promezhutochnykh ehnergij privela k sozdaniyu proekta i konstruktsii kriticheskoj sborki. Yadernyj proekt ehtoj sborki podcherkivaet vazhnost' dannykh poperechnykh sechenij i teoreticheskoj interpretatsii ehksperimental'nykh rezul'tatov, imeyushchikh otnoshenie k usovershenstvovannom u nadteplovomu torievomu reaktoru. Tochnost' analiticheskikh metodov byla podtverzhdena pri analize ehksperimental'nykh rezul'tatov, poluchennykh na reaktore nulevoj moshchnosti ZPR-III. Provodyatsya sravneniya trekh konfiguratsij teploperedachi s ispol'zovaniem udvoennogo vremeni v kachestve optimal'nogo parametra. EHffekt proizvodstva izotopa protaktiniya-233 i urana pri otnositel'no poleznom ispol'zovanii nejtronov, vozmozhnye koehffitsienty vosproizvodstva i kharakteristiki vygoraniya otsenivayutsya v svyazi s netochnostyami v yadernykh poperechnykh secheniyakh. (author)

  19. ANALISIS SENSITIVITAS TURBULENSI ALIRAN PADA KANAL BAHAN BAKAR PWR BERBASIS CFD

    Directory of Open Access Journals (Sweden)

    Endiah Puji Hastuti

    2015-04-01

    Full Text Available Turbulensi aliran pendingin pada proses perpindahan panas berfungsi untuk meningkatkan nilai koefisien perpindahan panas, tidak terkecuali aliran dalam kanal bahan bakar. Program CFD (CFD=computational fluid dynamics, FLUENT adalah program komputasi berbasis elemen hingga (finite element yang mampu memprediksi dan menganalisis fenomena dinamika aliran fluida secara teliti. Program perhitungan CFD dipilih dalam penelitian ini karena selain akurat juga dapat memberikan visualisasi dengan baik. Penelitian ini bertujuan untuk memahami karakteristika perpindahan panas, massa dan momentum dari dinding rod bahan bakar ke pendingin secara visual, pada medan temperatur, medan tekanan, dan medan energi kinetika pendingin, sebagai fungsi dinamika aliran di dalam kanal, pada kondisi tunak dan transien. Analisis dinamika aliran pada kanal bahan bakar PWR berbasis CFD dilakukan dengan menggunakan sampel data reaktor PWR dengan daya 1000 MWe dengan susunan bahan bakar 17x17. Untuk menguji sensitivitas persamaan aliran yang sesuai dengan model aliran turbulen pada kanal bahan bakar dilakukan pemodelan dengan menggunakan persamaan k-omega (Ƙ-ω, k-epsilon (Ƙ-ε, dan Reynold stress model (RSM. Pada analisis sensitivitas aliran turbulen di dalam kanal digunakan model mesh hexahedral dengan memilih tiga geometri sel yang masing masing berukuran 0,5 mm; 0,2 mm dan 0,15 mm. Hasil analisis menunjukkan bahwa pada analisis kondisi tunak (steady state, terdapat hasil yang mirip pada model turbulen Ƙ-ε standard dan Ƙ-ω standard. Pengujian terhadap kriteria Dittus Boelter untuk bilangan Nusselt menunjukkan bahwa model Reynold stress model (RSM direkomendasikan. Analisis sensitivitas terhadap geometri mesh antara sel yang berukuran 0,5 mm, 0,2 mm dan 0,15 mm, menunjukkan bahwa geometri sel sebesar 0,5 mm telah mencukupi. Aliran turbulen berkembang penuh telah tercapai pada model LES dan DES, meskipun hanya dalam waktu singkat (3 s, model LES memerlukan waktu komputasi

  20. Moessbauer Effect Study of the Isomeric De-Excitation in Sn{sup 119m}; Etude, par Effet Moessbauer, de la Desexcitation Isomerique dans {sup 119m}Sn; 0418 0417 0423 0427 0415 041d 0418 0415 041c 0415 0421 0421 0411 0410 0423 0415 0420 0421 041a 041e 0413 041e 042d 0424 0424 0415 041a 0422 0410 041f 0420 0418 0418 0417 041e 041c 0415 0420 041d 041e 041c 0421 041d 042f 0422 0418 0418 0412 041e 0417 0411 0423 0416 0414 0415 041d 0418 042f 0412 041e 041b 041e 0412 0415 -119{sup m}; Estudio de la Desexcitacion Isomerica del {sup 119m}Sn Mediante el Efecto Moessbauer

    Energy Technology Data Exchange (ETDEWEB)

    Herber, R. H.; Stoeckler, H. A. [School of Chemistry, Rutgers, State University, New Brunswick, NJ (United States)

    1965-04-15

    aproximacion no relativistica) la energia de retroceso ha de ser inferior a Tilde-Operator 0,3 keV ( Tilde-Operator 7 kcal/mol), valor que es considerablemente menor que las energias de enlace quimico que entran en juego. (author) [Russian] Kak i v sluchae K-zahvata i gamma-izluchenija ego raspada kobal'ta-57 himicheskie posledstvija izomernogo raspada olova-119{sup m} mogut byt' izucheny po spektroskopicheskomu metodu Messbauera. Snjatie vozbuzhdenija s 245-dnevnogo izotopa v sostojanii i/2 soprovozhdaetsja jemissiej kaskadnogo gamma-izluchenija v 65,3kjev(M4) i 23,8kjev(M1-E2). Po metodu Mess''auera mogut byt' izucheny tri glavnyh himicheskih jeffekta jadernyh prevrashhenij, a imenno: a) posledstvija reakcii Sn{sup 118} (n, {gamma})Sn{sup 119m} i jeffekty soputstvujushhego gamma-izluchenija pri obluchenii v reaktore; '') dejstvie otdachi, soprovozhdajushhej raspad M4 i s) dejstvie vnutrennej konversii raspada M4. Jeti jeffekty byli izucheny na dvuokisi olova, metallicheskom (serom) olove i na tetrafenile olova, kazhdyj iz kotoryh byl markirovan izotopom olova-119{sup m}. V Sn{sup 119m}O{sub 2} ni vozdejstvie gamma-radiacii pri obluchenii v reaktore, ni jemissija (i vnutrennjaja konversija) radiacii v 65,3 kjev ne privodjat k zametnym izmenenijam formy ili polozhenija linii rezonansa. Jeto nabljudenie rezko otlichaetsja ot togo, chto proishodit v analogichnyh uslovijah pri raspade (K-jeahvate i gamma-izluchenii) Co{sup 57} O. V oboih sluchajah izvestny dva sostojanija okislenija metallicheskoj okisi [sootvetstvenno Fe (II), Fe(III) i Sn (I), Sn(IV)], no tol'ko v sluchae Co{sup 57} O, v rezul'tate predshestvovavshego jadernogo prevrashhenija, nabljudalos' nalichie oboih sostojanij. Analogichnye otricatel'nye rezul'taty poluchajutsja dlja metallicheskogo (serogo) slova, chto soglasuetsja s nabljudenijami nad kobal'tom-57, rassejannym v metallicheskom zheleze i ispol'zovannom v kachestve Messbauerovskogo istochnika. Odnako, kak i v sluchae mechennogo kobal'tom-57

  1. Containment of Radioactive Waste for Sea Disposal and Fisheries Off the Canadian Pacific Coast; La Mise en Recipients des Dechets Radioactifs en Vue de leur Elimination dans la Mer et la Protection des Pecheries Operant au Large de la Cote Canadienne de l'Ocean Pacifique; 0423 0414 0414 ; Confinamiento de Desechos Radiactivos para su Evacuacion en el Mar, en Relacion con las Pesquerias de la Costa Canadiense del Pacifico

    Energy Technology Data Exchange (ETDEWEB)

    Waldichuk, Michael [Fisheries Research Board of Canada Biological Station, Nanaimo, BC (Canada)

    1960-07-01

    radiacii sostojat glavnym obrazom iz zagrjaznennyh laboratornyh othodov, bolee krupnyh chastej oborudovanija, trupov zhivotnyh posle biologicheskih jeksperimentov s mechennymi atomami, aktivnyh zhidkostej i nekotoryh materialov, obluchennyh v reaktore. Za period s 1946 goda po 1957 god s beregov Kalifornii bylo sbrosheno othodov s nizkim soderzhaniem radioaktivnyh veshhestv v kolichestve 16288 55-gallonnyh bochek. Glavnymi kriterijami dlja Soedinennyh Shtatov pri udalenii radioaktivnyh othodov v more javljajutsja: 1) tehnika bezopasnosti pri perevozke othodov ot ih istochnika do mesta ih udalenija i 2) sootvetstvujushhie uslovija pri pogruzhenii ih v more. Nikakih trebovaniij otnositel'no celostnosti kontejnerov i ih soderzhimogo po pogruzhenii na glubinu ne pred{sup j}avljaetsja. Daetsja opisanie kontejnerov, kotorymi pol'zujutsja ili kotorye ran'she predlagalis' dlja zakljuchenija v nih radioaktivnyh othodov, naprimer: za- paennye metallicheskie korobki, adsorbcija glinoj i vplavka v steklo. Dlja togo, chtoby sdelat' radioaktivnye othody bezvrednymi dlja ryb i drugih vodnyh organizmov, nadlezhit 1) libo izolirovat' ih ot sredy, 2) libo razbavit' radioaktivnye othody do urovnja dopustimoj koncentracii. Predlagaetsja dlja udalenija tverdyh othodov s nizkim i srednim urovnem radioaktivnosti primenjat' sfericheskij rezervuar, sootvetstvenno skonstruirovannyj, germeticheski zakuporennyj i sposobnyj vyderzhivat' vysokie davlenija, kak sredstvo izolirovanija othodov ot morskoj sredy. Radioaktivnye othody mogut okazat' vozdejstvie na rybu v morskoj srede cherez : 1) prjamoe obluchenie udalennymi radioaktivnymi materialami; 2) zaglatyvanie organizmov, sluzhashhih pishhej dlja ryb, soderzhashhih koncentrirovannye radioizotopy; 3) obluchenie vodoj, soderzhashhej radioaktivnye iony i chasticy; i 4) zarazhenie donnymi materialami, imejushhimi vysokoe soderzhanie osevshih na dno radioizotopov. Vyrazhaetsja mnenie, chto prezhde, chem mozhet byt' razresheno udalenie v more radioaktivnyh

  2. Dispersion Curves for Phonons in Diamond; Courbes de Dispersion des Phonons dans le Diamant; Krivye dispersii dlya fononov v almaze; Curvas de Dispersion de los Fonones en el Diamante

    Energy Technology Data Exchange (ETDEWEB)

    Warren, J. L.; Wenzel, R. G.; Yarnell, J. L. [Los Alamos Scientific Laboratory, Los Alamos, NM (United States)

    1965-04-15

    recien mencionados, la tienen menos en el caso del diamante. (author) [Russian] Trehosnyj difrakcionnyj spektrometr nejtronov na Losalamosskom reaktore ''Omega Vest'' ispol'zovalsja dlja izmerenija nekotoryh krivyh dispersii dlja rasprostranenija fononov v napravlenijah [100] i [111] v almaze . Vse iz merenija provodilis' pri komnatnoj temperature. Obrazec predstavljal soboj korichnevyj promyshlennyj almaz tipa Na vesom 242,8 karat (48,56 grammov). Izuchenie javlenij difrakcii nejtronov pokazalo, chto almaz sostoit iz bol'shogo edinichnogo kristalla s mozaichnym razbrosom v razmere M Degree-Sign polnoj shiriny pri polumaksimume, pljus dvumja nebol'shimi oblastjami, razorientirovannymi na 3 - 5 Degree-Sign . Spektrometr programmirovalsja dlja operacii ''postojannaja''. Vo vseh sluchajah jenergija padajushhih nejtronov byla postojannoj, i rassejannye nejtrony terjali jenergiju. Dostatochnaja intensivnost' pri jenergii bombardirovki, dostatochno bol'shoj dlja vozbuzhdenija opticheskih form almaza, byla dostignuta v rezul'tate ispol'zovanija otrazhenija (1122) berillievogo monohromatora. K trudnostjam provedenija jeksperimentov, pomimo trudnostej, voznikajushhih v svjazi s nalichiem ochen' vysokih chastot fononov v almaze, otnositsja vozniknovenie anomal'nyh pikov, chto zatrudnjaet raspoznavanie tochek poljarizacii. Schitaetsja, chto neopredelennosti v izmerennyh chastotah sostavljajut velichinu porjadka 2-3 procentov. Naibolee razitel'nym rezul'tatom jetih izmerenij javljaetsja podtverzhdenie predpolozhenij, osnovannyh na dannyh otnositel'no pogloshhenija infrakrasnoj oblasti spektra i otnositel'no udel'nogo tepla, chto almaz ne javljaetsja gomologom s kremniem i germaniem. Otsutstvie sootvetstvija projavljaetsja v osnovnom v povedenii poperechnyh akusticheskih vetvej, kotorye v sluchajah vyrazhenija ih v sootvetstvujushhih privedennyh edinicah po svoej chastote pochti v dva raza vyshe v almaze, chem v kremnii i germanii. Ostal'nye nabljudaemye krivye dispersii dlja almaza

  3. The Development of Materials for Application to Control Rod Systems in Graphite moderated Reactors; Mise au Point de Materiaux pour les Dispositifs de Controle a Barres, Utilbes dans les Reacteurs Ralentis au Graphite; Razrabotka materialov , primenyaemykh v sistemakh upravlyayushchikh sterzhnej v reaktorakh s grafitovym zamedlitelem; Perfeccionamiento de Materiales Aplicables a las Barras de Control en los Reactores Moderados por Grafito

    Energy Technology Data Exchange (ETDEWEB)

    Wade, G. E.; Kempf, F. J. [Hanford Atomic Products Operation, General Electric Company, Richland, WA (United States)

    1964-06-15

    utilizando cualquiera de los materiales mencionados. Los canales de las barras requieren a menudo un revestimiento para proteger el moderador de grafito que los circunda contra los impactos y el efecto de desgaste debidos a la insercion de las barras y para asegurar que el canal conserve la alineacion correcta. Tales revestimientos deben consistir en materiales capaces de soportar la abrasion y el impacto, dotados de gran resistencia mecanica, de reducida seccion eficaz y aptos para trabajar sin refrigeracion. Se ha ensayado con ese fin el grafito pirolitico puro y en forma de mezclas, el oxido de aluminio y el carburo de silicio. Los datos obtenidos acerca de los danos fisicos y de irradiacion indican que algunos de estos materiales se prestan para el revestimiento de los canales de las barras en los reactores. (author) [Russian] Materialy, primenjaemye v sistemah upravljajushhih i avarijnyh sterzhnej reaktorov s grafitovym zamedlitelem i teplonositelem v trubkah, mogut byt' podrazdeleny na dve kategorii: materialy dlja izgotovlenija upravljajushhih sterzhnej i materialy dlja izgotovlenija rubashek rabochih kanalov sterzhnej. Materialy dlja izgotovlenija upravljajushhih sterzhnej, naprimer bor ili gadolinij, mogut sostavljat' edinoe celoe s obolochkoj sterzhnja, kak pri ispol'zovanii boristoj nerzhavejushhej stali, primenjaemoj dlja izgotovlenija avarijnyh sterzhnej. Drugoj metod sostoit v zakljuchenii spechennogo bloka, soderzhashhego bor, naprimer B{sub 4}C -grafit ili B{sub 4}C -aljuminij, v metallicheskuju obolochku. Sterzhni poslednego tipa podhodjat dlja celej regulirovanija vvidu povyshennogo procenta soderzhanija bora. Ispytanija i opyt izgotovlenija pokazyvajut, chto pri ispol'zovanii jetih materialov mozhno skonstruirovat' razlichnye tipy udovletvoritel'nyh sterzhnej;. V kanalah sterzhnej v reaktore chasto trebujutsja rubashki dlja zashhity okruzhajushhego grafitovogo zamedlitelja ot vozdejstvija nagruzok pri vvedenii sterzhnja i iznosa i dlja podderzhanija

  4. The Use of the Moessbauer Effect in Investigating the Chemical Effects of Nuclear Transmutations in Oxygenous Compounds of Manganese and Tin; Utilisation de l'Effet Moessbauer pour l'Etude des Effets Chimiques des Transformations Nucleaires dans les Composes de Manganese et d'Etain; 0418 0421 0421 041b 0415 0414 ; Empleo del Efecto Moessbauer para Investigar las Transformaciones Nucleares de Compuestos Oxigenados de Estano y de Manganeso

    Energy Technology Data Exchange (ETDEWEB)

    Nesmejanov, An. N.; Babeshkin, A. M.; Kosev, N. P.; Bekker, A. A.; Lebedev, V. A. [Moskovskij Gosudarstvennyj Universitet Im. M.V. Lomonosova, Moskva, SSSR (Russian Federation)

    1965-04-15

    7} a 10{sup 8} neutrones/cm{sup 2} * s emitido por una fuente de polonio-berilio. Los atomos calientes de {sup 56}Mn que se forman pueden ser hepta y tetravalentes o presentar estados de oxidacion inferiores. A partir de las muestras (SnO y SnO{sub 2}), irradiadas en un reactor con un flujo de 2,2 * 10{sup 13} neutrones/cm{sup 2}- s durante 1800 h, prepararon directamente fuentes para investigar la absorcion por resonancia de los cuantos gamma en un aparato que funciona a velocidad constante. Como absorbentes utilizaron los correspondientes compuestos de estano. Los atomos de retroceso del {sup 119m}Sn solo se han observado en el estado tetravalente. Basandose en los datos obtenidos, los autores examinan el mecanismo de estabilizacion de los atomos de retroceso. (author) [Russian] Obychnye metody izuchenija himicheskogo sostojanija atomov otdachi (himicheskij analiz, hromatografija, jelektroforez), svjazannye s fazovymi perehodami, takzhe kak i otzhig, dajut vozmozhnost' lish' kosvenno sudit' o mehanizme stabilizacii gorjachih atomov. Pri analize metodom rezonansnogo pogloshhenija gamma-kvantov bez otdachi v obrazec ne vnosjatsja nikakie izmenenija, mogushhie povlijat' na formy stabilizacii dazhe, esli oni malo ustojchivy. V nastojashhbj rabote issledovalis' formy stabilizacii atomov otdachi marganca-56 v obluchennyh rastvorah i tverdyh obrazcah permanganata, a takzhe olova-119{sup m} v dvuokisi i okisi olova. Obrazcy permanganata obluchalis' v parafinovom bloke na polonij-berillievom istochnike s potokom nejtronov 10{sup 7} - 10{sup 8} n/sm{sup 2}/sek. Gorjachie atomy marganca-56 obrazujutsja v +7,+ 4 i nizshih valentnyh sostojanijah. Iz obrazcov ( SnO i SnO{sub 2}), obluchennyh na reaktore potokom 2,2 * 10{sup 13} n/cm{sup 2}/sek v techenie 1800 chasov, neposredstvenno gotovilis' istochniki dlja issledovanija rezonansnogo pogloshhenija gamma-kvantov na ustanovke, rabotajushhej v rezhime postojannoj skorosti. V kachestve poglotitelej ispol

  5. Initial Operating Experience with the ''NPD'' Reactor; Experience recueillie pendant les premiers mois de fonctionnement du reacteur NPD; Pervyj opyt po ehkspluatatsii reaktora NPD; Experiencia inicial de funcionamiento del reactor NPD

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, L. G. [Hydro-Electric Power Commission of Ontario, Toronto, Ontario (Canada)

    1963-10-15

    raboty na vysokoj moshchnosti v techenie shesti nedel' dal koehffitsient ispol'zovaniya na moshchnost', ravnyj 70% Uluchsheniya, kotorye byli dostignuty, uvelichili bezopasnost', povysili ehkspluatatsionnye kharakteristiki i prodemonstrirovali potentsial'nye metody snizheniya kapital'nykh zatrat dlya budushchikh stantsij. Tak, naprimer, v tselyakh uluchsheniya ehkspluatatsionnykh kachestv byli vidoizmeneny uplotneniya valov na nasosakh okhladitelya pervogo kontura; oborudovanie tipa kholodil'nikov, ispol'zovavsheesya dlya regeneratsii para, bylo zameneno pogloshchayushchimi kolonkami s tsel'yu umen'sheniya poteri para v tyazheloj vode. Ustanavlivayutsya takzhe ogranichiteli potoka vody v liniyakh dlya vzyatiya prob s tsel'yu umen'sheniya poteri tyazheloj vody v sluchayakh neispravnostej v soedineniyakh. V dekabre 1962 goda dve odnovremennye utechki iz mashiny, proizvodyashchej zagruzku topliva, priveli k neobychnomu obstoyatel'stvu, pri kotorom znachitel'noe kolichestvo goryachej tyazheloj vody pod vysokim davleniem popalo v kameru reaktora, gde proizoshlo neznachitel'noe umen'shenie ee izotopnoj chistoty, kotoraya zatem byla prevyshena i reaktor byl vnov' pushchen v kontse mesyatsa. Vo vremya avarii vse ustrojstva po bezopasnosti rabotali tochno i obespechili uderzhanie tyazheloj vody. (author)

  6. Recoil Processes of Cr{sup 51} in Mixed Inorganic Systems; Processus de Recul de {sup 51}Cr dans des Melanges Inorganique; 041f 0420 041e 0426 0415 0421 0421 042b 0421 042f 0414 0420 0410 041c 0418 041e 0422 0414 0410 0427 0418 0425 0420 041e 041c 0410 -51 0412 0421 041c 0415 III 0410 041d 041d 042b 0425 041d 0415 041e 0420 0413 0410 041d 0418 0427 0415 0421 041a 0418 0425 0421 0418 0421 0422 0415 041c 0410 0425 ; Procesos de Retroceso del {sup 51}Cr en Sistemas Inorganicos Mixtos

    Energy Technology Data Exchange (ETDEWEB)

    Veljkovic, S. R.; Milenkovic, S. M.; Ratkovic, M. R. [Boris Kidric Institute of Nuclear Sciences, Vinca, Yugoslavia (Serbia); Faculty Of Natural Sciences, Belgrade University, Belgrade, Yugoslavia (Serbia)

    1965-04-15

    smesjah hromatov ili solej hroma s neorganicheskimi okislami. V kachestve okislov ispol'zovalis' AI{sub 2}O{sub 3}, SiO{sub 2} i MgO. Termicheskie jeffekty kontrolirovalis' parallel'no dlja ocenki chisto teplovyh vozdejstvij v reaktore i dlja ocenki vozmozhnogo vnutrennego nagreva v mishenjah. V sisteme, gde hromaty byli osazhdeny na MgO, nabljudalos' sil'noe vosstanovlenie do hrom-iona. Pri temperature obluchenija nejtronnye jeffekty prevalirovali nad termicheskimi processami i tol'ko nagrevanie do temperatury vyshe 600 Degree-Sign S davalo podobnyj rezul'tat. Ion hroma vo vremja obluchenija ne izmenilsja. Uvelichenie koncentracii hromatov privodilo k horosho izvestnomu uderzhaniju Cr{sup 51}O{sub 4}{sup =}. Hromaty, adsorbirovannye na glinozeme, sohranjali formu okisi tak zhe, kak i v sisteme s SiO{sub 2}. Teplovye jeffekty imeli to zhe samoe napravlenie. Ion hroma, absorbirovannyj na jetih okisjah, vedet sebja razlichno. V sisteme s glinozemom otmechalos' sil'noe okislenie, kotoroe znachitel'no prevyshalo vklad teplovyh processov. Pri sravnenii jeffekta obluchenija i nagreva v sisteme s SiO{sub 2} obnaruzhena nebol'shaja raznica, hotja bylo otmecheno bol'shoe diffuzionnoe obednenie poverhnosti ionami hroma. Obshhaja osobennost' vseh sistem sostoit v ochen' nebol'shoj koncentracii soedinenij hroma. Himija tonkogo sloja, po-vidimomu, otlichaetsja ot obychnogo povedenija bol'shih mass hromatov v bol'shinstve sluchaev, kak v teplovyh processah, tak i v processe otdachi. Jeto mozhet oznachat', chto processy v ''gorjachih tochkah'', nahodjas' v zavisimosti ot vseh komponentov, mogut davat' himicheskie produkty, chuvstvitel'nye k harakteru misheni. Obobshhennaja kartina processov s jadrami otdachi v tverdyh soedinenijah, po-vidimomu, nuzhdaetsja v dopolnitel'nom rassmotrenii voprosa o materiale matric. (author)

  7. Wear studies in the shearing process by means of irradiated tools; Etudes d'usure dans les operations de cisaillement, au moyen d'outils irradies; Issledovaniya problemy iznosa v protsesse skalyvaniya posredstvom obluchennykh instrumentov; Estudios de desgaste en las operaciones de cizallamiento, realizados con ayuda de herramientas irradiadas

    Energy Technology Data Exchange (ETDEWEB)

    Sata, Toshio; Abe, Kunio; Nakajima, Kiyoshi [Institute of Physical and Chemical Research, Komagome, Bunkyo-Ku, Tokyo (Japan)

    1962-01-15

    del punzon y de la matriz al emplear lubricantes y laminas metalicas de distintos tipos, obteniendose los resultados siguientes: 1) Al comienzo de la operacion, los indices de desgaste del punzon y de la matriz son muy elevados, pero disminuyen rapidamente, para estabilizarse despues de haber recortado entre 400 y 500 piezas; 2) El desgaste del punzon supera en un 20 por ciento al de la matriz; 3) Los lubricantes con aditivos especiales para altas presiones (compuestos de cloro, fosforo o azufre) reducen el desgaste de las herramientas, mientras que los aceites minerales refinados ejercen poco efecto; 4) Cuanto mas duro es el metal trabajado, tanto mayor es el desgaste de las herramientas; el desgaste al punzonar acero inoxidable y acero al silicio es tres y seis veces superior, respectivamente, al observado en el caso del acero pobre en carbono. (author) [Russian] Iznos instrumenta v protsesse skalyvaniya metallicheskogo lista issledovalsya s pomoshch'yu probojnikov i stal'nykh puansonov s bol'shoj skorost'yu shtampovki, obluchennykh v yadernom reaktore. Kruglye diski diametrom 10 mm shtampuyutsya iz stal'nogo lista s nizkim soderzhaniem ugleroda tolshchinoj 0,5 mm, nerzhaveyushchej stali i kremnievoj stali, sukhim sposobom i so smazkoj. Posle shtampovki izmeryalas' radioaktivnost' diskov otverstij. Kogda radioaktivnyj puanson zamenyalsya neradioaktivnym, to na diskakh obnaruzhivalas' neznachitel'naya radioaktivnost' v to vremya, kak radioaktivnost' v otverstiyakh edva izmenyalas'. EHto pokazalo, chto iznos puansona mozhno opredelit' po radioaktivnosti diskov, a iznos probojnikov - po radioaktivnosti otverstij. S pomoshch'yu ehtogo metoda iznos probojnikov i puansonov proveryalsya pri ispol'zovanii razlichnykh vidov smazochnogo materiala i metallicheskogo lista i byli polucheny sleduyushchie rezul'taty: 1) V nachale shtampovki skorost' iznosa kak probojnika, tak i puansona ochen' bol'shaya, no ona bystro snizhaetsya i stanovitsya, nakonets, postoyannoj posle

  8. Major accident analyses for experimental zero-power fast reactor assemblies; Analyse des accidents graves pouvant survenir dans les reacteurs experimentaux a neutrons rapides de puissance zero; Analiz krupnoj avarii dlya ehksperimental'ny kh reaktornykh ustanovok nulevoj moshchnosti na bystrykh nejtronakh; Analisis de los accidentes graves que pueden producirse en los reactores experimentales rapidos de potencia cero

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, G.; Barts, E. W.; Kapil, S.; Tomabechi, K. [Argonne National Laboratory, Argonne, IL (United States)

    1962-03-15

    'shogo reaktora s aktivnoj zonoj obychnogo sostava. Byla issledovana men'shaya aktivnaya zona s vysokim pustotnym koehffitsientom, kak potentsial'no bolee opasnaya sistema. U ehtikh dvukh sistem obnaruzhen ochen' razlichnyj vremennoj rezhim. V sluchayakh ser'eznykh avarij na ustanovkakh nulevoj moshchnosti atomy U{sup 235}, kotorye raspredeleny v plastinakh obogashchennogo urana, ochen' bystro nagrevayutsya, togda kak ostal'naya chast' aktivnoj zony po sushchestvu ostaetsya kholodnoj, i takim obrazom gazoobraznyj U{sup 235} sozdaet raspredelennoe davlenie. V doklade budet dano opisanie primeneniya k gazu Van der Vaalsa koda AX-I nejtronnoj fiziki i gidrodinamiki. Drugim vazhnym izmeneniem uravneniya sostoyaniya, ispol'zovannogo v kode, yavlyaetsya primenenie uravneniya Mie-Grinejzena, vyvedennoe iz teorii tverdogo sostoyaniya. EHto izmenenie daet vozmozhnost' bolee udovletvoritel'n o vyrazit' chlen davleniya dlya aktivnykh zon razlichnogo sostava. Vvidu togo, chto plastiny U{sup 235} s vysokim obogashcheniem v ustanovke nulevoj moshchnosti nagrevayutsya gorazdo bystree, chem obednennye uranovye plastiny, vozmozhnost' polucheniya rezul'tiruyushcheg o polozhitel'nogo ehffekta Dopplera namnogo bol'she v ehksperimental'noj ustanovke, chem v reaktore-razmnozhitel e ehkvivalentnoj moshchnosti. EHtot risk byl issledovan v otnoshenii ryada vozmozhnykh ustanovok. Ehti raschety ukazyvayut na to, chto koehffitsient Dopplera ustanovki nulevoj moshchnosti ne priobretaet opasnogo znacheniya, poka ne budut sozdany sistemy krupnykh ehnergeticheskikh reaktorov-razmnozhitelej na oksidnom toplive s ochen' myagkimi spektrami ehnergii nejtronov. (author)

  9. Present Status of Nitrogen Fixation by Reactor Radiation; Etat Actuel des Recherches sur l'oxydation directe de l'azote sous irradiation dans des reacteurs; Sovremennoe sostoyani opytov po okisleniyu azota izlucheniem iz reaktorov; Estado actual de las investigaciones sobre fijacion del nitrogeno por irradiacion en reactores

    Energy Technology Data Exchange (ETDEWEB)

    Harteck, P; Dondes, S [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1960-07-15

    sistema funktsioniruet v nastoyashchee vremya v Brukkhejvenskom reaktore v zakrytom konture. Dayutsya dannye o vliyanii temperatury, davleniya, proportsii smesi i intensivnosti izlucheniya, chto mozhet byt' ispol'zovano dlya proektirovaniya budushchego khimiko-yadernogo reaktora. Sushchestvuyushchaya sistema rabotaet pri davlenii v 10 atmosfer i pri .temperature v 150''oC. Temperatura zavisit ot ehnergii rasshchepleniya chastits, osvobozhdaemoj v steklyannom volokne, i ot ognestojkosti zakrytogo kontura. V nastoyashchee vremya stroitsya drugoj zakrytyj kontur, kotoryj dolzhen rabotat' pri davlenii v 50-75 atmosfer i pri temperature v 600''oC. EHti zakrytye kontury dayut vozmozhnost' sudit' o parametrakh potochnykh sistem, vklyuchaya povedenie produktov rasshchepleniya, vypuskaemykh v gazovyj potok. Slozhnaya kinetika okisleniya azota opisyvaetsya na trekh stadiyakh : nachal'nye reaktsii v sisteme; reaktsii posle obrazovaniya nekotorogo kolichestva okislennogo azota i, nakonets, kinetika pri ustanovlenii ravnovesiya izlucheniya. Rassmatrivayuts ya usloviya obrazovaniya N{sub 2}0{sub 5}, N{sub 2}0{sub 4} i O{sub 3}, a takzhe ikh vliyanie na ves' protsess. (author)

  10. A Study of the Recoil Reactions of Three Isotopes of Ruthenium in Ruthenocene; Etude des Reactions d'Atomes de Recul de Trois Isotopes du Ruthenium dans le Ruthenocen; 0418 0417 0423 0427 0415 041d 0418 0415 0420 0415 0410 041a 0426 0418 0419 041e 0422 0414 0410 0427 0418 0422 0420 0415 0425 0418 0417 041e 0422 041e 041f 041e 0412 0420 0423 0422 0415 041d 0418 042f 0412 0420 0423 0422 0415 041d 041e 0421 0415 041d 0415 ; Estudio de las Reacciones de Retroceso de Tres Isotopos del Rutenio en el Rutenoceno

    Energy Technology Data Exchange (ETDEWEB)

    Harbottle, G.; Zahn, U. [Brookhaven National Laboratory, Upton, NY (United States)

    1965-04-15

    de la viscosidad del medio, se investigaron tambien soluciones al 2% de rutenoceno en benceno que contenian 2, 10 y 20% de poliestireno. Los resultados fueron muy semejantes a los obtenidos en el caso del benceno puro. Tambien se estudiaron soluciones congeladas al 2 y 0,2% de rutenoceno en benceno: en este caso, los valores de retencion se aproximaron a los obtenidos para los cristales de rutenoceno a la misma temperatura. Si se define el 'efecto isotopico' como [1-(retencion de {sup 97}Ru/retencion de {sup 103}Ru)J x 100 (expresado en tanto por ciento), el valor de dicho efecto correspondiente a las soluciones de rutenoceno es de 42% ot por termino medio, mientras que los cristales y soluciones congeladas presentan valores mas bajos, que oscilan entre 10 y 15%. La mayor retencion y el menor efecto isotopico quiza puedan explicarse por un mayor amortiguamiento del impulso de retroceso en una red cristalina que en un liquido, aunque este sea viscoso. Pero tambien han de tenerse en cuenta otros factores como la relajacion del estado de carga de los distintos isotopos del rutenio consecutiva a la desexcitacion del estado nuclear compuesto. Se discutiran otras posibles explicaciones de estos resultados. (author) [Russian] Reakcii otdachi treh izotopov rutenija izuchalis' v molekuljarnom ''pirozhkovom'' soedinenii rutenosena [(C{sub 5}H{sub 5}){sub 2}RuJ . Izotopami, kotorye byli proizvedeny posredstvom reakcij (p, {gamma}) v reaktore, byli Ru{sup 97} (2,9 dnja), Ru{sup 103} (40 dnej) i Ru{sup 105}) (4,45 chasa), i ih otnositel'naja aktivnost' opredeljalas' posredstvom mnogokanal'nogo analiza scintilljacionnyh spektrov. Kristally rutenosena bombardirovalis' pri razlichnyh temperaturah i raspolozhenijah, i nabljudaemoe uderzhanie sostavljalo dlja uslovij komnatnoj temperatury Ru{sup 97}, 9.5{+-}0.1; Ru{sup 103}, 10.7 {+-} 0.2; i Ru{sup 105} , 9.9{+-}0.2%. sootvetstvenno. Otzhig kristallov pri temperature 140 Degree-Sign C vyzyval nebol'shoe uvelichenie uderzhanija, odnako

  11. Criteria for Special Nuclear Materials Inventory and Control Procedures; Criteres a Suivre Pour Proceder a l'Inventaire des Matieres Nucleaires Speciales et aux Mesures de Controle; Kriterii dlya inventarizatsii spetsial'nykh yadernykh materialov i metody ucheta; Criterios a Que Deben Ajustarse los Procedimientos de Inventario y Control de los Materiales Nucleares Especiales

    Energy Technology Data Exchange (ETDEWEB)

    Kinderman, E. M.; Tarrice, R. R. [Stanford Research Institute, Menlo Park, CA (United States)

    1966-02-15

    bol'shinstvom promyshlennyh materialov, naprimer v SShA ceny na uran 90%-go obogashhenija i na uran 3%-go obogashhenija v vide shestiftoristogo urana i na tjazheluju vodu sostavljajut so otvetstvenno 10 808, 254 i 61,60 doll. SShA za 1 k g . Bolee togo, vo mnogih sluchajah jeti materialy, kak togo trebu- et ohrana zdorov'ja i tehnika bezopasnosti, nahodjatsja pod special'nym pravitel'stvennym kontrolem, ne svjazannym neposredstvenno s ih denezhnoj stoimost'ju. Nesmotrja na bol'shuju stoimost' jetih materialov, predusmatrivaetsja primenjat' ih v bol'shom kolichestve; napri- mer, v reaktore s vodnym zamedlitelem moshhnost'ju 500 mgvt budet ispol'zovano priblizi- tel'no 50 - 75 t materiala 3%-go obogashhenija, i, verojatno, vo v s em mire k 1980 godu moshhnost' reaktorov, nahodjashhihsja v jekspluatacii, budet jekvivalentno ravna moshhnosti priblizi- tel'no 200 - 300 reaktorov takogo razmera. Na osnovanii proshlogo opyta razrabotany special'nye metody i nalazhena praktika promyshlennogo ucheta nedorogostojashhih materialov v bol'shom kolichestve, naprimer ugol' ili zheleznaja ruda, i dorogostojashhih materialov v nebol'shom kolichestve, naprimer dragocen- nye metally . Pri pochti odinakovyh cenah special'nye jadernye materialy razlichajutsja po vidu i budut ispol'zovat'sja v kolichestvah, znachitel'no ''ol'shih po sravneniju s dragocennymi metallami. Hotja, verojatno, potrebujutsja special'nye metody ili sootvetstvuju- shhee izmenenie staryh metodov, nadlezhashhee ispol'zovanie mnogoobraznoj ustanovlennoj praktiki proverki i ucheta dolzhno okazat'sja dostatochnym v bol'shinstve sluchaev dlja dolzhnoj zashhity kapitalovlozhenij stran i otdel'nyh lic v proizvodstvo jetih dorogostojashhih materialov. Ustanavlivajutsja kriterii dlja ucheta materialov. Special'no rassmatrivaetsja vopros o, sootvetstvii razlichnyh metodov inventarnogo kontrolja, nachinaja ot sostavlenija ezhegodnyh balansov uch et a postuplenij i otpravok do podrobnoj ezhednevnoj fizicheskoj inventarnoj

  12. Design of a Chemical Processing Apparatus for Radioisotopes of Short Half-Life; Projet d'une Installation de Traitement Chimique de Radioelements a Courtes Periodes; ПРОЕКТ УСТАНОВКИ ДЛЯ ХИМИЧЕСКОЙ ОБРАБОТКИ КОРОТКОЖИВУЩИХ РАДИОЭЛЕМЕНТОВ; Proyecto para una Instalacion de Tratamiento Quimico de Radioelementos de Periodo Corto

    Energy Technology Data Exchange (ETDEWEB)

    Douis, M.; Valade, J. [Centre d' Etudes Nucleaires, Saclay (France)

    1963-03-15

    : Disolucion de los radioelementos destinados a fines diversos y preparacion del mercurio-197. (author) [Russian] Predstavilos' interesnym pokazat' radiohimikam, imejushhim v svoem rasporjazhenii nebol'shoj reaktor (naprimer, s potokom 10{sup 12}g n/cm{sup 2}/sek), vozmozhnosti poluchenija rjada radiojelementov minimal'nymi sredstvami. Jeta rabota sostoit iz treh osnovnyh chastej: 1. Vozmozhnye radiojelementy ; podrazdeljajutsja na dve kategorii: a) radiojelementy dlja primenenija v medicine, k kotorym otnosjatsja: natrij-24, kalij-42,brom-82, med' 64, mysh'jak-76, rtut'-197 i kolloidal'noe zoloto-190; b) radiojelementy dlja nauchnogo ili promyshlennogo primenenija, k kotorym, pomimo vysheukazannyh, otnosjatsja: sur'ma-122, mysh'jak-77, marganec-56, zoloto-196 (hloristoe). 2. Himicheskaja obrabotka; podrazdeljaetsja na dve gruppy: a) perevedenie v istinnye rastvory, chto trebuet znachitel'nogo kolichestva vremeni dlja rastvorenija v vode, razbavlennoj kislote na holodu, libo rastvorenija v koncentrirovannyh kislotah pri nagrevanii; v jetu gruppu vhodjat: natrij-24, kadij-42, brom-82, rtut'-197, sur'ma-122, marganec-56 i zoloto-196 (hloristoe); b) processy vydelenija ili slozhnyh prevrashhenij, kuda otnosjatsja metody vydelenija, osnovannye na jeffekte Scilarda-Chalmersa, reakcii (n, p), (p, {gamma}) s posledujushhim beta-raspadom idi obrazovaniem kolloidov; sjuda vhodjat: med'-64, mysh'jak-76, mysh'jak-77 i kolloidal'noe zoloto-198; 3. Zashhitnye kamery dlja predvaritel'noj obrabotki, gruppirujutsja soglasno ih svojstvam. Predlagaetsja ustanovka, sostojashhaja iz treh kamer dlinoj 2 i glubinoj v 1 m, kotorye soedineny mezhdu soboj konvejerom i igrajut rol' : 1-ja kamera: vvedenie kontejnerov, vyemka i prigotovlenie natrija-24, kalija-42 i broma-62; 2-ja kamera: prigotovlenie dvuh iz treh sledujushhih radiojelementov: med'-64, mysh'jak-76 i kolloidal'noe zoloto-198; 3-ja kamera: rastvorenie radiojelementov dlja razlichnogo primenenija i prigotovlenie rtuti-197. (author)

  13. Ternary Fission of U{sup 235} by Resonance Neutrons; Fission Ternaire de {sup 235}U par des Neutrons de Resonance; 0422 0420 041e 0419 041d 041e 0415 0414 0415 041b 0415 041d 0418 0415 0423 0420 0410 041d 0410 -235 041d 0410 0420 0415 0417 041e 041d 0410 041d 0421 041d 042b 0425 041d 0415 0419 0422 0420 041e 041d 0410 0425 ; Fision Ternaria del {sup 235}U por Neutrones de Resonancia

    Energy Technology Data Exchange (ETDEWEB)

    Kvitek, I.; Popov, Ju. P.; Rjabov, Ju. V. [Ob' edinennyj Institut Jadernyh Issledovanij, Dubna, SSSR (Russian Federation)

    1965-07-15

    }). Izmerenija provodilis' na impul'snom bystrom reaktore Ob'edinennogo instituta jadernyh issledovanij po metodu vremeni proleta. Ispol'zovalas' proletnaja baza dlinoj 100 m, chto obespechivalo razreshenie 0,6 mksek/m. Dlja registracii oskolkov delenija i legkoj dlinnoprobezhnoj chasticy ispol'zovalis' gazovye scintilljacionnye schetchiki, napolnennye ksenonom do davlenija 2 atm. Sloj oboga shennogo urana-235 tolshhinoj Tilde-Operator 2 mg/cm{sup 2} i ploshhad'ju-300 cm{sup 2} byl iajesen na aljuminievuju fol'gu tolshhinoj 20 mikron. V gazovom ob{sup e}me po odnu storonu fol'gi registrirovalis' scintilljacii ot oskolkov delenija, po druguju-ot legkih dlinnoprobezhnyh chastic. Dlja ocenki fonov (naprimer, sovpadenija impul'sa ot oskolka s impul'som ot {gamma}-kvanta delenija ili protona ot (p, r) reakcii na aljuminievoj fol'ge) provedeno izmerenie, v kotorom ob{sup e}m, registrirujushhij dlinnoprobezhnuju chasticu, jekranirovalsja dopolnitel'nym aljuminievym fil'trom tolshhinoj 1 mm. Poluchennye rezul'taty jeksperimenta ukazyvajut na otsutstvie sushhestvennyh kolebanij otnoshenija sechenij trojnogo i dvojnogo delenij dlja urana-235, otmechavshihsja drugimi avtorami. V izmerenijah ne obnaruzheno nereguljarnosti v otnoshenii sechenij v oblasti jenergij 0,1-0,2 jev. V doklade obsuzhdaetsja vozmozhnoe vlijanie reakcii (n, {alpha}) na rezul'taty, poluchaemye v opytah, ne ispol'zujushhih sovpadenij dlinnoprobezhnoj chasticy s oskolkom delenija. (author)

  14. Power Reactor Design at Zero Power; Etudes de Reacteurs de Puissance, au Moyen de Machines de Puissance Zero; Konstruktsiya ehnergeticheskogo reaktora nulevoj moshchnosti; Diseno de Reactores Generadores con Ayuda de Reactores de Potencia Nula

    Energy Technology Data Exchange (ETDEWEB)

    Redman, W. C.; Plumlee, K. E.; Baird, Q. L. [Argonne National Laboratory, Argonne, IL (United States)

    1964-02-15

    obsuzhdenii programm ob{sup j}asnjajutsja obstojatel'stva, vlijajushhie na vybor jeksponen- cial'nyh i chisto kriticheskih sborok, maketov ustanovok nulevoj moshhnosti i jeksperimen- tov na samom reaktore dlja poluchenija neobhodimyh dannyh, a takzhe ta rol', kotoruju igrajut vspomogatel'nye analiticheskie raboty. Na konkretnyh primerah pokazyvaetsja, kakie mo- gut byt' polucheny dannye do nachala raboty reaktora na moshhnosti. Jeti dannye vkljuchajut ostatochnuju reaktivnost' vykljuchennogo reaktora, zapas reaktivnosti dlja jekspluatacionnyh nuzhd, temperaturnye kojefficienty, jeffektivnost' regulirujushhih i avarijnyh sterzhnej, kinetiku reaktora, harakteristiki proizvodstva jenergii, trebovanija v otnoshenii puskovogo istochnika i chuvstvitel'nosti priborov, trebovanija v otnoshenii zashhity i balans nejtro- nov . V obzore nedavnih jeksperimentov v reaktorah nulevoj moshhnosti vyjavljaetsja bol'shaja rol', kotoruju igrajut v poslednee vremja jeksponencial'nye i kriticheskie sistemy v vypol- nenii Argonskoj laboratoriej zadach po razrabotke konstrukcij reaktorov. Ih rol' v budu- shhem vidna iz kratkogo izlozhenija osushhestvljaemyh i zaplanirovannyh programm dlja sem'i dejstvujushhih reaktorov nulevoj moshhnosti Argonskoj laboratorii i ozhidaemogo popolnenija. (author)

  15. Interesting Developments in UO{sub 2} Technology; Progres interessants dans la technologie du bioxyde d'uranium; Interesnye usovershenstvovaniya tekhnologii UO{sub 2}; Recientes progresos en la tecnologia del UO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Robertson, J. A.L. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)

    1963-11-15

    vydelenie iz UO{sub 2} gazov, yavlyayushchikhsya produktami deleniya. V chastnosti, uvelichenie oblucheniya s 10{sup 15} do 10{sup 18} delenij/cm{sup 2} mozhet snizit' ochevidnye skorosti diffuzii dlya ksenona v UO{sub 2} pri posleduyushchikh obzhigakh na koehffitsient 10{sup 3}. Gaz, po-vidimomu, uderzhivaetsya v mel'chajshikh lovushkakh, chast' iz kotorykh sushchestvuet v iskhodnom materiale, a chast' obrazuetsya v rezul'tate radiatsionnogo povrezhdeniya. Tshchatel'nyj analiz pokazal sushchestvovanie medlennoj utechki iz lovushek, chto, veroyatno, ob''yasnyaetsya ogranichennoj rastvorimost'yu ksenona v UO{sub 2}. Vozmozhnost' osushchestvleniya izmerenij v reaktore otkryvaet novuyu fazu eshche bolee vazhnykh ehksperimentov. Oni pokazhut, imeyutsya li kakie-libo potentsial'nye ehkonomicheskie preimushchestva v novykh formakh topliva. V to zhe vremya budut prodolzhat'sya nastojchivye razrabotki spechenoj UO{sub 2} v prostoj geometrii sterzhnya. (author)

  16. Shippingport Atomic Power Station Operating Experience, Developments and Future Plans; La centrale nucleaire de Shippingport, experience de son fonctionnement et plans pour l'avenir; Shippingportskaya atomnaya ehlektrostantsij, opyt ehkspluatatsii, usovershenstvovaniya i plany na budushchee; Central nucleoelectrica de Shippingport; experiencia adquirida con su explotacton y programa de desarrollo

    Energy Technology Data Exchange (ETDEWEB)

    Feinroth, H. [Division of Reactor Development, United States Atomic Energy Commission, Washington, DC (United States); Oldham, G. M. [Shippingport Atomic Power Station, Duquesne Light Company, Pittsburgh, PA (United States); Stiefel, J. T. [Bettis Atomic Power.Labora Tory, Westinghouse Electric Corporation, Pittsburgh, PA (United States)

    1963-10-15

    elevada densidad de potencia, el denominado cuerpo No. 2. Con una potencia nominal bruta de 150 MW(e) y una duracion equivalente a 10 000 h de funcionamiento a plena potencia, el cuerpo No. 2 dara una produccion de energia 5,5 mayor que la del cuerpo No. 1 y su potencia especifica sera el doble de la de este ultimo. Se describen las caracteristicas de diseflo del cuerpo No. 2 y se resumen los principales adelantos en materia de fisica de reactores, metalurgia, transmision de calor, circulacion de fluidos, y elaboracion de elementos combustibles. Por ultimo, se describen los planes para la descontaminacion de la central nuclear y para introducir en la misma las modificaciones exigidas por la instalacion del cuerpo No. 2, de mayor potencia. (author) [Russian] Daetsya otsenka pyati godam ehkspluatatsii i ispytanie Shippingportskoj ehlektrostantsii, a takzhe rassmatrivayutsya poslednie tekhnicheskie usovershenstvovaniya i programma na budushchee. Ehta programma napravlena na usovershenstvovanie osnovnoj tekhnologii reaktorov na obychnoj vode s tem; chtoby sozdat' osnovu dlya umen'sheniya v budushchem stoimosti yadernoj ehlektroehnergii. Shippingportskaya reaktornaya ustanovka, ehkspluatirovavshayasya pyat' let, priznana godnoj dlya legkogo podklyucheniya k sisteme ehnergosnabzheniya v kachestve ustanovki dlya bazovoj ili pikovoj nagruzok. Rabota komponentov ustanovki byla nadezhnoj. Ne voznikalo problem, svyazannykh s zagryazneniem ili udaleniem otkhodov. Dostup k komponentam pervichnoj sistemy okhlazhdeniya dlya ikh obsluzhivaniya 'byl khoroshim, chto pokazyvaet tselostnost' toplivnykh ehlementov. Kazhdaya iz trekh operatsij po zamene topliva v reaktore s momenta ego puska trebovala vse men'she i men'she vremeni. Nedavno byla ustanovlena tret'ya zapal'naya sborka, na chto potrebovalos' 32 rabochikh dnya, t.e. okolo odnoj chetverti vremeni, potrachennogo na pervuyu zamenu topliva. Opisany ofitsial'nye trebovaniya v otnoshenii podgotovki personala, pis

  17. Some Applications of Short-Lived Radioisotopes in the Study of Metals; Applications Diverses des Radioelements de Courte Periode dans l'Etude des Metaux; РАЗЛИЧНЫЕ ПРИМЕНЕНИЯ КОРОТКОЖИВУЩИХ РАДИОЭЛЕМЕНТОВ ПРИ ИССЛЕДОВАНИЯХ МЕТАЛЛОВ; Aplicaciones de los Radioelementos de Periodo Corto en el Estudio de los Metales

    Energy Technology Data Exchange (ETDEWEB)

    Kohn, A. [Institut de Recherches de la Siderurgie, Saint-Germain-en-Laye (France)

    1963-03-15

    despues de la colada. En muchos de estos lingotes, se registraron corrientes de conveccion de intensidad suficiente para distribuir el oro por gran parte del volumen del lingote una hora despues de terminada la colada. (author) [Russian] V jetom doklade izlagaetsja nekotorye tipichnye vidy primenenija korotkozhivushhih radiojelementov pri issledovanii metallov. Razrabotan prostoj metod aktivacionnogo analiza dlja opredelenija soderhanija lantana v razlichnyh markah stali, k kotoroj vo vremja plavki dobavljaetsja neznachitel'noe kolichestvo neochishhennoj smesi redkozemel'nyh jelementov. Issledovalos' povedenie mysh'jaka v period okislenija zheleza s cel'ju izuchit', kak vedut sebja neznachitel'nye kolichestva soderzhashhegosja v zheleze mysh'jaka v jetot period. S pomoshh'ju avtoradiografii bylo ustanovleno prezhde vsego znachitel'noe obogashhenie mysh'jakom poverhnosti razdela metall-okis'. V rezul'tate primenenija metoda, zakljuchajushhegosja v aktivirovanii okislennyh obrazcov v jader- nom reaktore i rastvorenii obrazujushhihsja sloev tolshhinoj v neskol'ko mikron, predstavilos' vozmozhnym izuchit' dannoe javlenie v kachestvennom otnoshenii. Udalos' proverit', chto mysh'jak koncentriruetsja v metalle rjadom s poverhnost'ju razdela, gde ego koncentracija mozhet v 30 - 40 raz prevyshat' pervonachal'noe soderzhanie primesi. S pomoshh'ju avtoradiografii issledovalsja prirost metallicheskih kristallov, s uchetom togo, chto tverdoe telo bolee bedno legirujushhimi jelementami po sravneniju s rasplavom, iz kotorogo ono obrazuetsja. Jeto issledovanie kasalos' medlenno ohlazhdaemyh splavov alljuminija i medi, kotorye zatem poluchali zakalku v processe zatverdenija. Issledovanie pozvolilo vyjavit' konfiguraciju metallicheskih kristallov na razlichnyh stadijah ih prirosta i ustanovit' zavisimost' processa zatverdenija ot izmenenija temperatury splava, izmerjaemoj s pomoshh'ju termopary. Izuchenie konvektivnyh tokov v bol'shih slitkah dlja pokovki imelo cel'ju podtverdit' nalichie

  18. Review of Development Status of Nuclear Superheat; Expose sur l'etat actuel des travaux concernant la surchauffe nucleaire; Obzor razrabotki voprosa o yadernykh peregrevatelyakh; Estudio de los progresos realizados en niateria de sobrecalentamiento nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Imhoff, D. H.; Pennington, R. T. [General Electric Company, San Jose, CA (United States)

    1963-10-15

    nuclear utilizado en el sobrecalentador, evaluaciones de la corrosion uniforme y localizada, tanto en el interior como en el exterior del reactor, resultados de experimentos criticos de sobrecalentamiento termico, y de las comprobaciones experimentales de transmision de calor; por ultimo, examen sucinto de las ventajas economicas que, segun los estudios, presentan los reactores de sobrecalentador separado, los reactores de sobrecalentador integrado y los reactores con sobrecalentamiento de espectro mixto. b) Breve descripcion del programa ESADA-VESR de desarrollo de combustibles para sobrecalentamiento nuclear, patrocinado por la Comision de Energia Atomica. Examen de los trabajos de investigacion, del diseflo de los elementos combustibles de la primera carga del sobrecalentador, de la gama de variables experimentales y de los resultados previstos para el programa trienal de desarrollo del combustible. (author) [Russian] Nachinaya s 1959 goda kompaniya ''Dzheneral Ehlektrik'' aktivno zanimalas' provedeniem raboty v oblasti yadernogo peregreva na reaktorakh s zamedlitelem iz obychnoj vody. Za ehtot period v SSHA vpervye byl proizveden yadernyj peregretyj par v khode provedeniya usovershenstvovannogo demonstratsionnogo ehksperimenta s peregrevom cha ustanovke SADE. Ehtot proekt finansirovalsya kompaniej. Nyneshnee sostoyanie s yadernym peregrevom podrazdelyaetsya na dve glavnye kategorii. Pervaya yavlyaetsya opisaniem trekh osnovnykh ustanovok po oblucheniyu peregretogo topliva, ispol'zuemykh kompaniej ''Dzheneral ehlektrik'', i vtoraya - opisaniem dvukh glavnykh razrabotannykh programm deyatel'nosti, vmeste s obzorom po sostoyaniyu na segodnyashnij den' znachitel'nykh rezul'tatov razvitiya v oblasti peregreva. 1. Glavnye usovershenstvovannye ustanovki: a) Daehtsya kratkoe opisanie ehksperimenta (SADE) na Vallesitosskom reaktore s kipyashchej vodoj (VBWR), privodyatsya tablitsy ehkspluatatsionnykh uslovij, dannye o toplivnykh ehlementakh, obluchennykh v period mezhdu maem

  19. Void Reactivity Effects in the Second Charge of the Halden Boiling Water Reactor; Effets Cavitaires dans la Deuxieme Charge du Reacteur a Eau Lourde Bouillante de Halden (HBWR); Ehffekty pustotnoj reaktivnosti vo vtoroj zag HBWR; Effectos de Cavitacion en la Segunda Carga del Reactor de Agua Pesada Hirviente de Halden (HBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Lunde, J. E. [OECD Halden Reactor Project (Norway)

    1964-02-15

    temperaturas elevadas. La concordancia es menos satisfactoria para valores intermedios del volumen relativo de los vacios. El efecto de reactividad se calcula macroscopicamente mediante una teoria de las perturbaciones. (author) [Russian] Jeffekt reaktivnosti pustot, vyzvannyh kipeniem v kanalah dlja teplonositelja vo vtoroj zagruzke Haldejskogo kipjashhego tjazhelovodnogo reaktora (HBWR) byl izmeren kak v jeksperimentah nulevoj moshhnosti s imitaciej pustot, tak i v uslovijah fakticheskoj moshhnosti. Jeksperimenty s imitirovannymi pustotami sostojali v izmerenii jeffekta reaktivnosti pustotnyh kolonn, vvodimyh v tonkostennye trubki na razlichnuju glubinu. Trubki byli ustanovleny v raznyh polozhenijah mezhdu prodol'nymi rebrami v odinochnoj sborke, sostojashhej iz semi sterzhnej, prakticheski identichnoj normal'nym toplivnym jelementam vtoroj zagruzki. Jetot jeksperiment pozvolil izuchit' zavisimost' reaktivnosti ot pustogo ob{sup e}ma, a takzhe zavisimost' reaktivnosti ot polozhenija puzyr'kov para v kanale dlja teplonositelja. Jeksperiment byl vypolnen na norvezhskoj ustanovke nulevoj moshhnosti NORA s aktivnoj zonoj iz 36 jelementov vtoroj zagruzki i s geometriej reshetki, analogichnoj geometrii reshetki reaktora HBWR. Temperaturnaja zavisimost' pustotnogo jeffekta byla izuchena na jeksperimental'noj ustanovke nulevoj moshhnosti, imevshej aktivnuju zonu Iz 100 toplivnyh jelementov reaktora HBWR. V odinochnom toplivnom jelemente uroven' vody v kanale dlja teplonositelja snizhalsja do razlichnyh glubin i vlijanie na reaktivnost' jetogo otklonenija ot normal'nyh uslovij izmerjalos' pri razlichnyh temperaturah a intervale temperatur 50 -220eS. Vlijanie pustotnoj reaktivnosti na moshhnost' bylo izmereno na reaktore HBWR kak funkcija jadernoj jenergii pri razlichnyh temperaturah zamedlitelja v diapazone 150 -230e S pri moshhnostjah do 16mgvt (pri samoj vysokoj temperature). Kojefficient pustotnoj reaktivnosti na moshhnosti javljaetsja vazhnym pokazatelem pri opredelenij

  20. Thermal Vibrations of Beta-Brass and the Order-Disorder Transition; Vibrations Thermiques dans le Laiton Beta et Transformation Ordre-Desordre; Teplovye kolebaniya beta-latuni i perekhod iz uporyadochennogo sostoyaniya v razuporyadochennoe; Vibraciones Termicas del Laton Beta y Transicion Orden-Desorden

    Energy Technology Data Exchange (ETDEWEB)

    Dolling, G.; Gilat, G. [Chalk River Nuclear Laboratories, Chalk River, ON (Canada)

    1965-04-15

    reciproco, que presentan acusados aumentos de la anchura energetica a la temperatura de transicion. Estos efectos no se han podido explicar aun de manera logica. (author) [Russian] Nabljudenija za obychnymi formami kolebanija uporjadochennogo splava medi i cinka ({beta}-latun') pri 296 Degree-Sign K provodilis' s pomoshh'ju kogerentnogo odno fononnogo rassejanija medlennyh nejtronov na monokristallicheskih obrazcah. Dlja izmerenija chastot kolebanij, rasprostranjajushhihsja vdol' vysokosimmetrichnyh napravlenij [00{zeta}], [{zeta}{zeta}0], [{zeta}{zeta}{zeta}{zeta}] i [ Vulgar-Fraction-One-Half Vulgar-Fraction-One-Half {zeta}] ispol'zovalsja trehosnyj kristallicheskij spektrometr na reaktore NRU. Krivye dispersii napominajut krivye dispersii prostogo ob{sup e}mno-centrirovannogo kubicheskogo kristalla, naprimer Na, za iskljucheniem togo, chto pojavljajutsja nekotorye vyrozhdenija, v osnovnom vvidu razlichija mezhdu vtorym blizhajshim sosedom Si-Si i silami Zn-Zn. Naprimer, my nahodim dve razlichnye formy volnovogo vektora (0.5, 0.5, 0.5) (v obratnyh edinicah reshetki), chastoty kotoryh sootvetstvenno sostavljajut (4,21 {+-}0,06) i (4,93 10) * 10{sup 12} gerc. Kratko upominajutsja modeli mezhatomnyh sil, kotorye predstavljajut udovletvoritel'noe opisanie rezul'tatov, poluchennyh pri 296 Degree-Sign K. Nekotorye normal'nye formy kolebanij izucheny pri povyshennyh temperaturah, osobenno v oblasti perehoda iz uporjadochennogo sostojanija v neuporjadochennoe pri temperature priblizitel'no 727 Degree-Sign K. V obshhej strukture krivyh dispersij ne otmechaetsja znachitel'nyh izmenenij v rezul'tate ischeznovenija porjadka dal'nego rasstojanija pri jetoj temperature, hotja razlichnye ''rasshheplenija'', nabljudaemye pri 296 Degree-Sign K, rasplyvajutsja v bolee ili menee postojannye ''polosy'' chastot. Pri povyshenii temperatury chastoty v celom umen'shajutsja, a jenergeticheskie shiriny uvelichivajutsja. Jeti izmenenija proishodjat plavno, za iskljucheniem dvuh prodol

  1. Neutron Investigation of Magnon Spectrum in Haematite; Etude du Spectre de Magnons dans l'Hematite, au Moyen des Neutrons; Nejtronnoe issledovanie spektra magnona v gematite; Estudio, por Metodos Neutronicos, del Espectro de Magnones en la Hematita

    Energy Technology Data Exchange (ETDEWEB)

    Dimitrijevic, Z.; Rzany, H.; Todorovic, J.; Wanic, A. [Institute for Nuclear Physics Cracow (Poland)

    1965-04-15

    -Sign y 126 Degree-Sign , respectivamente. Un acuerdo satisfactorio entre la teoria y los datos experimentales disponibles se tiene cuando SJ{sub 1} = SJ{sub 2} = 5,1 meV. Ademas de la rama magnonica acustica, los calculos han revelado la existencia de una rama optica, pero el estudio de esta no pudo llevarse a cabo por el metodo de difraccion neutronica. (author) [Russian] byli provedeny v Vincha na reaktore RA s ispol'zovaniem kristallicheskogo nejtronnogo spektrometra. Rassejanie monohromaticheskih nejtronov ({lambda} = 1.314A) proizvodilos' na krupnom monokristalle gematita ({alpha}-Fe{sub 2}O{sub 3}). Issledovalis' uglovye raspredelenija neuprugo rassejannyh nejtronov. Dlja rjada razlichnyh nepravil'nyh uglov kristalla AO byla izmerena shirina G puchka rassejannyh nejtronov (tak nazyvaemogo konusa rassejanija). Jetot konus rassejanija byl pripisan javleniju poverhnostnogo rassejanija magnona vokrug tochki obratnoj reshetki [1,1,1] . Dlja rjada skorostej magnona byla rasschitana i sravnena s jeksperimental'nymi tochkami zavisimost' G ot nepravil'nogo ugla. Ustanovleno, chto velichina skorosti v napravlenii [111] ravna 25,5 11,0 km/sek. Byla obnaruzhena strukturnaja anizotropija dispersionnogo sootnoshenija magnona. Bylo najdeno, chto skorost' javljaetsja bolee vysokoj v napravlenijah rasprostranenija parallel'nyh osi [111] i chto kachestvenno ona horosho soglasuetsja s bolee rannimi izmerenijami (Riste i dr.) dajushhimi v = 38km/sek. Obnaruzhit' v akusticheskoj oblasti jenergii magnona sushhestvovanie razryva jenergeticheskoj krivoj Eg ne udalos'. Bylo rasschitano, chto velichina E dolzhna byt' nizhe 1 Mjev. Pri pomoshhi formalizma Uollase byli rasschitany dispersionnye sootnoshenija magnona v gematite. Pri jetom ishodili ije predpolozhenija, chto sushhestvujut dva ne stremjashhihsja k nulju integrala obmena J{sub 1} i J{sub 2}. J{sub 1} i J{sub 2} oznachajut svjaz' sverhobmennoj jenergii mezhdu spinami sosednih ionov zheleza, svjazannyh ionom kisloroda. Ugly svjazi

  2. The Use of Research Reactors and Short-Lived Isotopes in the Study of Nuclear-Reactor Fuel Materials; Emploi de Reacteurs de Recherche et de Radioisotopes de Courte Periode dans l'Etude des Combustibles pour Reacteurs Nucleaires; ИСПОЛЬЗОВАНИЕ ИССЛЕДОВАТЕЛЬСКИХ РЕАКТОРОВ И КОРОТКОЖИВУЩИХ ИЗОТОПОВ ПРИ ИЗУЧЕНИИ ТОПЛИВНЫХ МАТЕРИАЛОВ ДЛЯ ЯДЕРННХ РЕАКТОРОВ; Empleo de Reactores de Investigacion y de Isotopos de Periodo Corto en el Estudio de Combustibles Nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Elleman, T. S.; Townley, C. W.; Sunderman, D. N. [Battelle Memorial Institute, Columbus, OH (United States)

    1963-03-15

    tipos de materiales sin que sea preciso emplear grandes reactores de ensayo ni recintos blindados para manipular muestras irradiadas de elevada actividad. (author) [Russian] Issledovatel'skij reaktor mozhet byt' ispol'zovan dlja izuchenija podvizhnosti produktov delenija v prototipah jadernyh toplivnyh materialov, tak kak on pozvoljaet vosproizvodit' vneshnie uslovija analogichnye tem, v kotoryh okazyvajutsja toplivnye materialy v uslovijah normal'noj jekspluatacii. Vmeste s tem on pozvoljaet tochno kontrolirovat' uslovija jeksperimenta i obespechivaet mnogoobrazie jeksperimental'nyh konstrukcij. Izmeneniem uslovij obluchenija i putem kolichestvennogo opredelenija vydeljajushhihsja iz obrazca produktov delenija s korotkim poluperiodom raspada mozhno ustanovit' mehanizm vysvobozhdenija produktov delenija i ego svjaz' s fizicheskimi i himicheskimi svojstvami obrazca topliva i produktov delenija. Pomimo jetogo, mogut byt' polucheny poleznye svedenija otnositel'no obshhego kolichestva vydeljaemoj radioaktivnosti i pred- polagaemogogo sroka raboty toplivnyh jelementov. Obrazcy obychno obluchajutsja v podogrevaemyh kapsulah s dvojnymi stenkami, pogruzhaemyh a bassejn reaktora ili v reaktornye kanaly dlja obluchenija, a vydeljaemye letuchie produkty delenija uvlekajutsja iz kapsuly struej gaza. Vvidu togo, chto sootnoshenie mezhdu skorost'ju vysvobozhdenija i poluperiodom zhizni radioizotopa javljajutsja vazhnym pokazatelem mehanizma vydelenija, sobirajutsja i analizirujutsja gazy delenija kripton i ksenon s poluperiodami zhizni ot 1,7 sek do 5,3 dnej. Korotkozhivushhie redkie gazy (kripton-89, kripton-91, kripton-92, ksenon-137, ksenon-138, ksenon-139, ksenon-140 i ksenon-141) opredeljajutsja putem sbora neletuchih radioaktivnyh dochernih produktov na zarjazhennom jelektrode dlja posledujushhego radiohimicheskogo analiza, togda kak gazoobraznye produkty delenija s bolee dlitel'nym poluperiodom zhizni (kripton-85, kripton-87, kripton-88, jod-131, ksenon-133 i ksenon-135

  3. Pathogenesis of Intrapulmonary Haemorrhage in Dogs Exposed to Pulsed Fission-Spectrum Neutrons; Pathogenese de l'Hemorragie Intrapulmonaire chez des Chiens Exposes a des Neutrons de Fission Pulses; Patogenez vnutrilegochnykh krovoizliyanij u sobak pri obluchenii impul'snymi nejtronami spektra deleniya; Patogenesis de la Hemorragia Intrapulmonar en Perros Expuestos a Neutrones de Fision Pulsados

    Energy Technology Data Exchange (ETDEWEB)

    Jones, R. K.; Engel, R. E.; Godden, W. R. [Kirtland AFB, New Mexico (United States)

    1964-05-15

    cambios se generalizan aun mas a los nueve dias, observandose tambien zonas focales de extravasacion peribronquial de la sangre alrededor de los dos bronquios primarios. La autopsia al decimotercer dia de la irradiacion revelo una extensa hemorragia perivascular que comprende las arterias pulmonares gruesas y finas. La sangre aparece dentro del tejido conjuntivo adventicio y de los espacios linfaticos periarteriales. La hinchazon del endotelio es muy pronunciada en los vasos afectados, pero la zona media de estos no presenta modificaciones esenciales. Tambien se manifiesta una hemorragia peribronquial contigua a las ramas afectadas de la arteria pulmonar. La patogenesis de estas alteraciones guarda al parecer una relaciuen con una lesion endotelial primaria producida por los neutrones de fision pulsados. Es probable que tambien actue como factor complementario una trombocitopenia, ya que se comprobo que el numero de megacariocitos medulares es inversamente proporcional a la gravedad de la hemorragia intrapulmonar. (author) [Russian] Predydushhie issledovanija v dannoj laboratorii po tkanevym povrezhdenijam u sobak, vyzyvaemym rentgenovskim oblucheniem s pikovym naprjazheniem 250 kv i impul'snymi nejtronami spektra delenija pokazali obshhee shodstvo reakcii. Odnako u zhivotnyh, obluchennyh nejtronami spektra delenija, chashhe v o z nikajut peribronhial'nye i perivaskuljarnye legochnye krovoizlijanija. Sdelana popytka opredelit' patogenez jetih porazhenij putem obluchenija pjati gonchih sobak dozoj v 400 rad bystryh nejtronov na reaktore tipa GODIVA bez otrazhatelja i serijnogo zabivanija zhivotnyh cherez opredelennye intervaly. Ch erez +5 dnej posle obluchenija mozhno bylo videt' znachitel'nye patologicheskie iz menenija, sostojashhie iz redkih melkih perifericheskih l egochnyh krovoizlijanij. K 9-mu dnju oni stanovilis' bolee mnogochislennymi. Naibolee porazitel'nye izmenenija, nabljudavshiesja k + 13-mu dnju, sostojali iz obshirnyh perivaskuljarnyh krovoizlijanij vokrug

  4. Short-Lived Isotopes Used as Tracers in Industry (with Special Reference to Swedish Industry); Emploi de Radioindicateurs de Courte Periode a l'Echelle Industrielie dans les Usines Suedoises; ПРИМЕНЕНИЕ КОРОТКОЖИВУЩИХ ИЗОТОПОВ В ПРОМЫШЛЕННОСТИ В КАЧЕСТВЕ МЕЧЕНЫХ АТОМОВ; Isotopos de Periodo Corto Utilizados Como Indicadores en la Industria Sueca

    Energy Technology Data Exchange (ETDEWEB)

    Erwall, L. G.; Forsberg, H. G.; Ljunggren, K. [Isotoptekniska Laboratoriet, Stockholm (Sweden)

    1963-03-15

    ; - obnaruzhenie utechek. Chtoby svesti k minimumu zagrjaznenija okonchatel'nyh produktov pri provedenii mnogih iz jetih issledovanij, bylo sushhestvenno vazhno ispol'zovat' korotkozhivushhie mechenye atomy. V bol'shinstve sluchaev primenjalis' metody fizicheskogo mechenija s vozmozhnost'ju vybora samyh raznoobraznyh mechenyh atomov. Tem samym predstavljaetsja vozmozhnym vybirat' mechenye atomy s podhodjashhim poluperiodom raspada i s nuzhnymi harakteristikami izluchenija. Esli ispol'zuemyj radioizotop obladaet bol'shim secheniem aktivacii nejtronov, chto prisushhe perechislennym vyshe izotopam, to issledovatel'skij reaktor mozhet dat' dostatochnyj potok nejtronov dlja proizvodstva aktivnostej indikatorov porjadka do neskol'kih kjuri pri uslovii, chto ne trebuetsja dlja nih znachitel'noj udel'noj aktivnosti. V Shvecii v 1960 godu byl obrazovan ob{sup e}dinennyj promyshlennyj institut i laboratorija izotopnyh metodov dlja provedenija issledovanij i rabot po razvitiju, a takzhe dlja okazanija pomoshhi promyshlennym predprijatijam posredstvom predostavlenija konsul'tacij, personala i neobhodimyh priborov po promyshlennomu primeneniju radioizotopov. (author)

  5. Evaluation of Fluorine-18 as a Scanning Agent for Intracranial Tumours; Emploi du Fluor-18 en Scintigraphy des Tumeurs Intracraniennes; Otsenka Ftora-18 kak skenniruyushchego agenta dlya vnutricherepnykh opukholej; Examen de las Caracteristicas del Fluor-18 como Agente de Exploracion de los Tumores Intracraneales

    Energy Technology Data Exchange (ETDEWEB)

    Entzian, W.; Aronow, S.; Soloway, A. H.; Sweet, W. H. [Massachusetts General Hospital, Boston, MA (United States)

    1964-10-15

    una curva de depuracion sanguinea con dos constantes de tiempo. La primera era muy breve y la segunda mas larga. Esta secuencia es tipica de los agentes de exploracion eficaces. (author) [Russian] V kachestve verojatnogo agenta dlja lokalizacii mozgovyh opuholej izuchalsja ftor-18, izluchatel' pozitronov s periodom poluraspada 110 min. Ftor-18 poluchali v jadernom reaktore putem obluchenija Li{sub 2}CO{sub 3} (Li{sup 6} (n, {alpha}) H{sup 3}, O{sup 16}(H{sup 3}, n) F{sup 18}) teplovymi nejtronami. Tkanevye issledovanija na myshah s peresazhennoj podkozhno jependimomoj pokazalo, chto mechenyj ftorid natrija (NaF{sup 18}) sozdaet izbytochnye koncentracii v kostjah i pojetomu ne udovletvorjaet v kachestve skennirujushhego agenta dlja mozga. Odnako pri primenenii mechenogo ftorborata kalija (KBF{sup 18}{sub 4})obnaruzheno izbiratel'noe pogloshhenie ego opuhol'ju po sravneniju s mozgom, prichem velichina pogloshhenija izmenjalas' v diapazone ot 8 do 11. Kak u myshej, tak i u koshek ne obnaruzheno izbytochnoj koncentracii v drugih organah. Izuchenie toksichnosti provodilos' na myshah i krolikah pri koncentracii 300mg/kg i 120 mg/kg sootvetstvenno. Pri nejtralizacii rastvora do urovnja fiziologicheskogo pH ne nabljudalos' nikakih neblagoprijatnyh jeffektov. Himicheskie kolichestva, neobhodimye dlja skennirovanija pacientov, sostavljali menee 0,5 mg/kg vesa tela. Sravnitel'nye klinicheskie issledovanija u 10 pacientov, kazhdyj iz kotoryh podvergalsja skennirovaniju s KBF{sup 18}{sub 4} drugimi izotopnymi veshhestvami, podtverdili vyvody Ashkenazi i dr. o tom, chto jeto soedinenie javljaetsja poleznym dlja opredelenija lokalizacii mozgovyh opuholej. Vnutrivennoe vvedenie osushhestvljalos' optimal'no za 30 minut do skennirovanija pri doze 15 mkkjuri/kg vesa tela. U odnogo pacienta pri operacii byli vzjaty obrazcy tkanej opuholi i zdorovoj tkani mozga. Kojefficient pogloshhenija sostavljal 3,5. Hotja jeto sootnoshenie nizhe, chem u myshej, no jeta velichina prakticheski prigodna

  6. Biological Effects of Thermal Neutrons and the B{sup 10}(n, {alpha}) Li{sup 7} reaction; Effets Biologiques des Neutrons Thermiques et la Reaction {sup 10}B(n, {alpha}){sup 7}Li; Biologicheskoe dejstvie teplovykh nejtronov i reaktsiya B{sup 10}(n, {alpha}) Li{sup 7}; Efectos Biologicos de los Neutrones Termicos y la Reaccion {sup 10}B(n, {alpha}){sup 7}Li

    Energy Technology Data Exchange (ETDEWEB)

    Archambeau, J. O.; Alcober, V.; Calvo, W. G.; Brenneis, H. [Medical Research Center, Brookhaven National Laboratory, Upton, NY (United States)

    1964-05-15

    captura de neutrones termicos en el tejido de recubrimiento. La irradiacion de la cabeza de un perro con un flujo nvt de 1,4 x 10{sup 14} n/cm{sup 2} provoca depilacion, eritema y escamacion humeda, con una depresion hematologica concomitante. Sin embargo, a los 25 o 30 dias se observa un restablecimiento de la medula osea y una normalizacion d el estado de la piel. Sometiendo a los animales a una irradiacion con un nvt igual a 5 x 10{sup 13} n/cm{sup 2} 30 min despues de administrar por via intravenosa 35 mg/kg de boro-10, se observa una epidermitis necrotizante, un edema del cuero cabelludo y una conjuntivitis. El cerebro presenta hemorragias capilares y estasis, con lesiones de las neuronas y astrocitos, asi como alteraciones del endotelio capilar. A continuacion, se observa una pronunciada disminucion del numero de plaquetas, que agrava las alteraciones locales. Los animales mueren a raiz de hemorragias y/o lesiones cerebrales entre el quinto y el noveno dias a contar de la irradiacion. Los efectos se atribuyen tanto a los rayos gamma como a las particulas alfa producidas por la captura de neutrones por el boro {sup 10}B(n, {alpha}){sup 7}Li. La irradiacion de la piel del cerdo con un tlujo nvt de 5 x 10{sup 12} n/cm{sup 2} no produce alteraciones histologicas. En cambio, irradiando la p iel con el mismo nvt despues de una inyeccion intravenosa de 35 mg/kg de boro-10, se provoca una radioepidermitis de tipo clasico, que se cura a los 36 o 40 dias. Fraccionando el nvt total en 8 aplicaciones en un periodo de 12 dias no se reduce la gravedad de la reaccion como ocurre en el caso de los rayos gamma. Este efecto se atribuye principalmente a la irradiacion por particulas a lfa formadas a raiz de la captura de neutrones termicos por el boro-10. (author) [Russian] Obluchenie zhivotnyh teplovymi nejtronami, poluchaemymi v medicinskom issledovatel'skom reaktore, privodit k tkanevym izmenenijam, javljajushhimsja rezul'tatom korpuskuljarnogo i gamma

  7. Reactor Physics Development for Advanced Gas-Cooled Reactors; Recherches en Physique des Reacteurs, pour des Reacteurs Perfectionnes Refroidis par un Gaz; Razrabotka metodov v oblasti reaktornoj fiziki dlya usovershenstvovannogo reaktora s gazovym okhlazhdeniem; Progresos de la Fisica de los Reactores de Tipo Avanzado Refrigerados por Gas

    Energy Technology Data Exchange (ETDEWEB)

    Moore, J. [United Kingdom Atomic Energy Authority (United Kingdom)

    1964-04-15

    reshetki dlja reaktora AGR i proverki teoreticheskih metodov, razrabotannyh dlja geterogennyh aktivnyh zon reaktorov, ispol'zovalis' kriticheskaja ustanovka APEX i reaktor nulevoj moshhnosti HERO s obychnymi raspolozhenijami reshetok , i kombinacijami izmenjajushhih rabotu reaktora jelementov, naprimer regulirujushhih sterzhnej. Teoreticheskie metody, razrabotannye i primenjavshiesja do nastojashhego vremeni, izvestny kak ''getrekontrol'' i FTD2. Jeksperimenty imeli cel'ju podrobno proverit' osobennosti jetih metodov, i dlja opredelenija soglasovannogo mezhdu soboj rjada konstant reshetki, sootvetstvujushhih rezul'tatam jeksperimentov, byli proanalizirovany rezul'taty izmerenij, vypolnennyh na rjade ''reaktornyh'' aktivnyh zon razlichnogo razmera v ustanovkah APEX i HERO. Jeti chisto jempiricheskie konstanty byli zatem ispol'zovany v metodah getrekontrol' i FTD2 dlja uspeshnogo planirovanija vvoda v jekspluataciju i vybora vida nagruzki dlja Uindskejlskogo AGR. Daetsja ssylka na jeksperimental'nye metody, kotorye byli provereny ilj special'no razrabotany dlja reshenija vstretivshihsja problem. Osobyj interes predstavljajut metody, ispol'zovavshiesja dlja izmerenija jeffektov reaktivnosti v reaktorah APEX, HERO i AGR i dlja opredelenija dannyh tonkoj struktury i raspredelenija jenergij v slozhnyh toplivnyh sborkah. Osushhestvljaemye v nastojashhee vremja teoreticheskie raboty skoncentrirovany, glavnym obrazom, na razrabotke al'ternativnogo metoda v otnoshenii ''getrekontrolja'' i FTD2 dlja rascheta aktivnyh zon reaktora posle znachitel'nogo vygoranija topliva. Na jetom zhe budut skoncentrirovany raboty i v budushhem. Zadachej programmy jeksperimentov na ustanovke HERO javljaetsja ispytanie jetih metodov na slozhnyh aktivnyh zonah, vkljuchaja aktivnye zony s toplivom, proizvodjashhim plutonij. Dopolnitel'nye dannye o vlijanii plutonija bydut polucheny blagodarja jekspluatacii reaktora AGR i fizicheskim izmerenijam obluchennogo topliva. (author)

  8. MASURCA, a Fast-Neutron Critical Mock-Up: Operation and Uses; MASURCA. Maquette Critique a Neutrons Rapides. Description Fonctionnelle et Obiectifs; ''MAZURKA'' - kriticheskaya model' na bystrykh nejtronakh. funktsional'noe opisanie i tseli; Descripcion Funcional y Objetivos de la Maqueta Critica de Neutrones Rapidos 'MASURCA '

    Energy Technology Data Exchange (ETDEWEB)

    Schmitt, A. P.; Storrer, F.; Vendryes, G. [Association CEA-EURATOM, Cadarache (France); Tavernier, G.; Van Dievoet, J. [Societe Belgo-Nucleaire, Bruxelles (Belgium)

    1964-02-15

    i vysotoj 102 mm. Oni razmeshheny v trubah kvadratnogo sechenija s vneshnej storonoj 106 mm i dlinoj porjadka 4 m. Jeti truby nahodjatsja v pod- veshennom sostojanii rjadom drug s drugom. Central'naja chast' podveshennoj plity mozhet prinimat' men'shie truby, chto pozvoljaet sozdavat' aktivnye zony nebol'shih razmerov. Special'nyj podogretyj kontur takzhe mozhet peremeshhat'sja v central'noj chasti ustanovki, chto daet vozmozhnost' izmerjat' kojefficient Dopplera. V doklade daetsja funkcional'noe opisanie ustanovki. Ukazyvaetsja na vozmozhnost' osushhestvlenija modelirovanija aktivnyh zon bystryh reaktorov na okislah (s pomoshh'ju jelementov iz okisi zheleza dlja vvedenija kisloroda) i na karbidah (s pomoshh'ju grafitovyh jele- mentov). Obsuzhdaetsja metod smeshannyh zagruzok, ko.toryj neobhodimo budet primenjat' vvidu opredelennogo kolichestva plutonija, ispol'zuemogo v takih opytah v nastojashhem desja- tiletii, i metodov, kotorye budut primenjat'sja dlja opredelenija osnovnyh nejtronnyh pa- rametrov. V zakljuchenie ukazyvaetsja mesto, kotoroe zanimaet jetot kriticheskij maket sredi os- novnyh fizicheskih issledovanij, svjazannyh s razrabotkoj reaktorov na bystryh nejtronah. V chastnosti, ispol'zovannye jeksperimental'nye sredstva vkljuchali reaktor-istochnik ''Garmonija'', prednaznachennyj dlja razrabotki,metodov izmerenija i dlja obespechenija jeksponenci- al'nyh opytov, a takzhe uskoriteli v kachestve istochnika staticheskih i dinamicheskih nejtronov dlja obespechenija podkriticheskih opytov na bystryh nejtronah. (author)

  9. Establishment of the Processes of Absorption and Diffusion of Systemic Insecticides in Populus Euramericana Dode Guinier ''Robusta''; L'etablissement des processus d'absorption et diffusion des insecticides systemiques au Populus x Euramericana Dode Guinier ''Robista; Opredelenie protsessov pogloshcheniya i diffuzii somaticheskikh insektitsidov u Populus x Euramerican a Dode Guinier ''Robusta''; Determinacion de los procesos de absorcion y difusion de los insecticidas sistemicos en el Populus x Euramericana Dope Guinier ''Robusta''

    Energy Technology Data Exchange (ETDEWEB)

    Catrina, I.; Popa, A.; Constantinesco, V.; Constantinesco, O.; Constantinesco, El.; Hulula, C. [Institut de Recherches Forestieres de Bucarest, Bucharest (Romania)

    1963-09-15

    observado que en el suelo queda gran cantidad de insecticida, los autores consideran que en la lucha contra los insectos que atacan a las especies lenosas es preferible verter en el suelo las soluciones de insecticidas de accion indirecta. (author) [Russian] Fosforoorganicheskie insektitsidy s somaticheskimi svojstvami, svyazannymi s ikh sposobnost'yu pronikat' v sok rastenij, vozdejstvuyut na nasekomykh, kotorye provodyat chast' svoej zhizni libo v kambial'noj zone mekhdu drevesinoj i koroj, libo v drevesine. Do poyavleniya ehtikh insektitsidov khimicheskaya bor'ba protiv ksilofagovykh nasekomykh, nakhodyashchikhsya v stvole rastenij, byla pochti nevozmozhnoj. Dlya izucheniya mekhanizma pogloshcheniya, diffuzii i lokalizatsii somaticheskikh insektitsidov u topolya i ivy, kotorye chasto podvergayutsya napadeniyu ksilofagovykh nasekomykh, byli provedeny issledovaniya s primeneniem mechenogo ''Dipterex'' na topole Robusta R-20. Mechenie insektitsida bylo proizvedeno v reaktore s ispol'zovaniem v kachestve misheni ''Dipterex'' v poroshke ( 1,5 g) pri potoke v {Phi} = 10{sup 11} n/cm{sup 2} sek i pri temperature v 30 - 40{sup o}C. Ehto prodolzhalos' do polucheniya absolyutnoj aktivnosti misheni {approx_equal} 1 mc. Chast' opytov provodilas' v laboratoriyakh, gde ispol'zovalis' rasteniya, vyrashchennye v vegetatsionnykh sosudakh. Drugaya chast' opytov provodilas' na pochve, v pitomnikakh, s ispol'zovaniem odnoletnikh i dvukhletnikh rastenij. V techenie 1 - 2 mesyatsev posle vvedeniya rastvorov mechenogo insektitsida rasteniya podvergalis' radiometricheskomu analizu. Pri provedenii opytov v laboratornykh usloviyakh insektitsid nakaplivalsya v znachitel'nykh kolichestvakh v listve i v neznachitel'nykh kolichestvakh - v drevesine. Pri provedenii opytov kak v laboratornykh usloviyakh, tak i na pochve insektitsid nakaplivalsya v bol'shem protsentnom otnoshenii v listve. Vmeste s tem otmechaetsya znachitel'nyj rost nakoplenij insektitsida v vetvyakh i v drevesine stvola, osobenno v

  10. Application of Sodium-24 to Flow-Rate Measurements and Leak Detection; Applications du Sodium-24 a des Mesures de Debits et de Recherches de Fuites; ИСПОЛЬЗОВАНИЕ НАТРИЯ-24 ПРИ ИЗМЕРЕНИЯХ ПОТОКОВ И ИССЛЕДОВАНИЯХ УТЕЧКИ; Aplicaciones del Sodio-24 a la Medicion de Caudales y Deteccion de Escapes

    Energy Technology Data Exchange (ETDEWEB)

    Guizerix, J.; Cornuet, R. [Centre d' Etudes Nucleaires, Grenoble (France)

    1963-03-15

    de transito de la onda de actividad entre dos detectores de centelleo colimados permitio determinar el caudal con una precision relativa del 2%, aproximadamente. (author) [Russian] Avtory opisyvajut nekotorye opyty, v kotoryh v kachestve indikatora ispol'zovalsja natrij-24 s cel'ju iskljuchenija problemy dezaktivacii i povtornogo vvedenija indikatorov. Obluchenie natrija bylo proizvedeno v bassejnovom reaktore 'Meljuzin' v Issledovatel'skom jadernom centre v Grenoble (Frakcija). S cel'ju issledovanija utechki v sisteme central'nogo otoplenija vodjanaja sistema metilas' karbonatom natrija-24. Posle ochishhenija i promyvki sistemy byla sostavlena tshhatel'naja shema issleduemogo uchast-. ka, ogranichivajushhaja zonu issledovanija. S pomoshh'ju kollimirovannyh priborov issledovalas' pochva, tochnye krivye ijeoaktivnosti issledovalis' dlja poluchenija razlichnyh dannyh. Obuslovlennoe utechkoj prevrashhenie v zavisimosti ot vremeni issleduemoj zony radioaktivnosti pozvolilo vydelit' ego na foke issleduemyh zon v rajone nepodvizhnyh trub ili v mestah sil'nogo pogloshhenija indikatora. Issledovalas' utechka na germeticheski zakrytom svincovom bake. Jetot bak pomeshhalsja mezhdu dvumja jarusami, pod nastilom rezervuara s vodoj, i byl prednaznachen dlja sbora vody, kotoraja mogla pojavit'sja v rezul'tate vozmozhnoj utechki v gidravlicheskih ustanovkah, raspolozhennyh v verhnem jaruse. V bak vvodilas' voda, mechennaja izotopom natrija-24. Nalozhenie radioaktivnyh pjaten v nizhnej i verhnej chastjah utechki, a takzhe slabyj vyhod potoka, ocenivaemogo v 1 sm{sup 3}/min, vyzvali izvestnuju trudnost' v issledovanii, i tol'ko chetkoe mestopolozhenie i rost radioaktivnogo pjatna pozvolili pri pomoshhi krivyh ijeoaktivnosti za 15 chasov opredelit' polozhenie utechki. V celjah opredelenija vozmozhnoj utechki porjadka 1 l/chas, kotoraja imeet mesto mezhdu sistemoj vodjanogo ohlazhdenija i sistemoj acetata celljulozy, potoki kotoryh izvestny i sostavljajut primerno 5 m3/chas

  11. The Formation of Polymeric Products in Reactions of Polyvalent Recoil Atoms; Formation de Polymeres lors de Reactions Provoquees par des Atomes de Recul Polyvalents; Obrazovanie polimernykh produktov pri reaktsiyakh polivalentnykh atomov otdachi; Formacion de Polimeros en las Reacciones de Atomos de Retroceso Polivalentes

    Energy Technology Data Exchange (ETDEWEB)

    Dzantiev, B. G.; Stukan, R. A.; Shvedchikov, A. P.; Shishkov, A. V. [Institut Himicheskoj Fiziki AN SSSR, Moskva, SSSR (Russian Federation)

    1965-04-15

    mechenyh polimernyh produktov v processe himicheskoj stabilizacii atomov otdachi sery-35 i ugleroda-14, poluchajushhihsja po jadernym reakcijam Cl{sup 35} /n, p/S{sup 35} i N{sup 14}/n, p/C{sup 14} v gazovoj i zhidkoj fazah. Mozhno predpolozhit', chto v processe stabilizacii gorjachie atomy ugleroda obrazujut metilenovye biradikaly, kotorye po svoej sposobnosti vstupat' v reakciju vo mnogom napominajut povedenie atomarnoj sery. Issledovanija provodilis' kak dlja parafinovyh (CH{sub 4}, C{sub 2}H{sub 6}), tak i dlja ciklicheskih (ciklogeksan, ciklogeksen, benzol) uglevodorod. Oblucheniju podvergalis' binarnye sistemy uglevodorod-datchik gorjachih atomov S{sup 35} i C{sup 14}. V kachestve poslednego ispol'zovalis' soedinenija CCI{sub 4}, HCl i ammiak. Obluchenie provodilos' na reaktore tipa IRT-1000 pri potoke teplovyh nejtronov 10{sup 11} - 10{sup 12} neJtron/cm{sup 2}sek. Pokazano, chto dlja razlichnyh soedinenij v zhidkoj faze do 60-90% sery-35 stabilizuetsja v forme polimera, vyhod kotorogo jekstremal'no zavisit ot sostava, prohodja cherez maksimum pri sootnoshenii komponentov, blizkom k jekvimolekuljarnomu. V gazovoj faze vyhod polimera sostavljaet 30 - 40% ot obshhej aktivnosti. Metodom radiohromatografii na bumage ustanovleno, chto mechenye polimernye produkty imejut slozhnyj sostav i predstavljajut soboj smes' dvuh kachestvenno otlichnyh tipov soedinenij, vyhod kotoryh po-raznomu menjaetsja v zavisimosti ot sootnoshenija komponentov. Uvelichenie vremeni obluchenija privodit k rostu vyhoda mechenogo polimera. V sluchae zhidkofaznoj sistemy C{sub 6}H{sub 12}-CCl{sub 4} molekuljarnyj ves S{sup 35}-soderzhashhego polimera, opredeljalsja metodom diffuzii iz kapilljara i okazalsja ravnym 5000 dlja polimera odnogo tipa i 500 - 1000 dlja drugogo. Obrazovanie vysokokipjashhego mechenogo produkta nabljudalos' takzhe pri obluchenii chistogo CCI{sub 4}. Analogichnye opyty provodilis' v sisteme jetilen -ammiak v gazovoj faze pri vysokom davlenii. Pokazano, chto v jetom sluchae

  12. Properties of Waste from Coal Gasification in Entrained Flow Reactors in the Aspect of Their Use in Mining Technology / Właściwości odpadów ze zgazowania węgla w reaktorach dyspersyjnych w aspekcie ich wykorzystania w technologiach górniczych

    Science.gov (United States)

    Pomykała, Radosław

    2013-06-01

    Most of the coal gasification plants based of one of the three main types of reactors: fixed bed, fluidized bed or entrained flow. In recent years, the last ones, which works as "slagging" reactors (due to the form of generated waste), are very popular among commercial installations. The article discusses the characteristics of the waste from coal gasification in entrained flow reactors, obtained from three foreign installations. The studies was conducted in terms of the possibilities of use these wastes in mining technologies, characteristic for Polish underground coal mines. The results were compared with the requirements of Polish Standards for the materials used in hydraulic backfill as well as suspension technology: solidification backfill and mixtures for gob caulking. Większość przemysłowych instalacji zgazowania węgla pracuje w oparciu o jeden z trzech głównych typów reaktorów: ze złożem stałym, dyspersyjny lub fluidalny. W zależności od rodzaju reaktora oraz szczegółowych rozwiązań instalacji, powstające uboczne produkty zgazowania mogą mieć różną postać. Zależy ona w dużej mierze od stosunku temperatury pracy reaktora do temperatury topnienia części mineralnych zawartych w paliwie, czyli do temperatury mięknienia i topnienia popiołu. W ostatnich latach bardzo dużą popularność wśród instalacji komercyjnych zdobywają reaktory dyspersyjne "żużlujące". W takich instalacjach żużel jest wychwytywany i studzony po wypłynięciu z reaktora. W niektórych przypadkach oprócz żużla powstaje jeszcze popiół lotny, wychwytywany w systemach odprowadzania spalin. Może być on pozyskiwany oddzielnie lub też zawracany do komory reaktora, gdzie ulega stopieniu. Wszystkie z analizowanych odpadów - trzy żużle oraz popiół pochodzą właśnie z tego typu instalacji. Tylko z jednej z nich pozyskano zarówno żużel jak i popiół, z pozostałych dwóch jedynie żużel. Odpady te powstały, jako uboczny produkt zgazowania w